23rd Symposium on Fusion Technology
20 - 24 September 2004 - Fondazione Cini, Venice, Italy
Book of Abstracts
Consorzio RFX – EURATOM - ENEA
1
PROGRAM
ORAL SESSIONS









O1-A
O1-B
O2-A
O2-B
O3-A
O3-B
O4-A
O4-B
Arazzi Hall
Barbantini Hall
Arazzi Hall
Barbantini Hall
Arazzi Hall
Barbantini Hall
Arazzi Hall
Barbantini Hall
Monday, 20th
Monday, 20th
Tuesday, 21th
Tuesday, 21th
Tuesday, 21th
Tuesday, 21th
Thursday, 23th
Thursday, 23th
14:00-15:30
14:00-15:30
09:00-09:30
09:00-09:30
14:00-15:30
14:00-15:30
14:00-15:30
14:00-15:30
Capriate Hall
Tipografia Hall
Capriate Hall
Tipografia Hall
Capriate Hall
Tipografia Hall
Capriate Hall
Tipografia Hall
Monday, 20th
Monday, 20th
Tuesday, 21th
Tuesday, 21th
Wednesday, 22th
Wednesday, 22th
Thursday, 23th
Thursday, 23th
16:00-18:00
16:00-18:00
16:00-18:00
16:00-18:00
11:30-13:30
11:30-13:30
16:00-18:00
16:00-18:00
POSTER SESSIONS








P1-C
P1-T
P2-C
P2-T
P3-C
P3-T
P4-C
P4-T
H-K-I
I
D
E
A-C
B
F
G-J
Paper Code
P 4 B – J – 491
ID NUMBER
TOPIC
HALL (A=Arazzi
B= Barbantini C =Capriate
SESSION
CATEGORY (P=Poster
Oral presentations are highlighted in red.
2
O=Oral)
T= Tipografia
INDEX
- A - Current and Next Step Devices ---------------------------------------------------------------------------------------------------------------------------------- 21
P3C-A-11
SELECTION OF DESIGN SOLUTIONS AND FABRICATION METHODS AND SUPPORTING R&D FOR PROCUREMENT
OF ITER VESSEL AND FW/BLANKET ---------------------------------------------------------------------------------------------------------- 21
P3C-A-16
THE PROTO-SPHERA LOAD ASSEMBLY ----------------------------------------------------------------------------------------------------- 22
P3C-A-72
O3B-A-72
OVERVIEW OF THE DIII–D PROGRAM AND CONSTRUCTION PLANS* ------------------------------------------------------- 23
P3C-A-90
COMMISSIONING AND PRELIMINARY OPERATION OF THE HL-2A TOKAMAK ------------------------------------------------ 24
P3C-A-144
PLASMA PHYSICS BASIS AND OPERATIONS OF FUSION-DRIVEN SUBCRITICAL SYSTEM --------------------------------- 25
P3C-A-184
THE WENDELSTEIN 7-X MECHANICAL STRUCTURE SUPPORT ELEMENTS: TESTS AND DESIGN ----------------------- 26
P3C-A-203
LEVITATION EXPERIMENTS OF A HIGH TEMPERATURE SUPERCONDUCTOR COIL IN THE INTERNAL COIL
DEVICE MINI-RT ------------------------------------------------------------------------------------------------------------------------------------- 27
P3C-A-245
EXPERIMENTAL STUDY OF WATER FLOW DISTRIBUTION INSIDE TWO-CHANNEL MODEL OF ITER VACUUM
VESSEL COOLING SYSTEM ---------------------------------------------------------------------------------------------------------------------- 28
P3C-A-262
MAGNUM-PSI, A PLASMA GENERATOR FOR PLASMA-SURFACE INTERACTION RESEARCH IN ITER-LIKE
CONDITIONS ------------------------------------------------------------------------------------------------------------------------------------------ 29
P3C-A-330
THE LASER MÉGAJOULE (LMJ) PROJECT DEDICATED TO INERTIAL CONFINEMENT FUSION :
DEVELOPMENT AND CONSTRUCTION STATUS. --------------------------------------------------------------------------------------- 30
O3B-A-330
P3C-A-449
O3B-A-449
JET ENGINEERING: PROGRESS AND PLANS -------------------------------------------------------------------------------------------- 31
P3C-A-468
THE JET-ENHANCED PERFORMANCE PROGRAM: MORE HEATING POWER AND DIAGNOSTIC CAPABILITIES IN
PREPARATION FOR ITER -------------------------------------------------------------------------------------------------------------------------- 32
P3C-A-518
NUCLEAR ANALYSES OF SOME KEY ASPECTS OF THE ITER DESIGN WITH MONTE CARLO CODES ------------------- 33
P3C-A-522
TRANSPORT, LOGISTICS AND PACKAGING OF ITER COMPONENTS --------------------------------------------------------------- 34
P3C-A-523
STUDIES FOR SITE PREPARATION FOR ITER CONSTRUCTION ---------------------------------------------------------------------- 35
P3C-A-531
READINESS OF CADARACHE FOR STARTING ITER CONSTRUCTION -------------------------------------------------------------- 36
3
- B - Plasma Heating and Current Drive.----------------------------------------------------------------------------------------------------------------------------- 37
P3T-B-19
THE ALCATOR C-MOD LOWER HYBRID CURRENT DRIVE EXPERIMENT TRANSMITTER ---------------------------------- 37
P3T-B-25
DESIGN OF AN ULTRA-BROADBAND SINGLE-DISK OUTPUT WINDOW FOR A FREQUENCY STEP-TUNABLE 1 MW
GYROTRON -------------------------------------------------------------------------------------------------------------------------------------------- 38
P3T-B-51
EXPERIMENTS ON A 170 GHZ COAXIAL CAVITY GYROTRON ----------------------------------------------------------------------- 39
P3T-B-78
THE UPGRADE OF THE DIII-D EC SYSTEM USING 120 GHZ ITER GYROTRONS ------------------------------------------------- 40
P3T-B-82
THE LHCD LAUNCHER FOR ALCATOR C-MOD – DESIGN, CONSTRUCTION, CALIBRATION AND TESTING* --------- 41
P3T-B-94
DESIGN AND OPERATION OF THE WENDELSTEIN 7-X ECRH HIGH VOLTAGE POWER SUPPLIES ----------------------- 42
P3T-B-113
THERMAL ANALYSIS AND OHMIC LOSS ESTIMATION OF POLARIZER FOR ITER ECCD SYSTEM ----------------------- 43
P3T-B-152
TESTS OF LOAD-TOLERANT EXTERNAL CONJUGATE-T MATCHING SYSTEM FOR A2 ICRF ANTENNA AT JET ---- 44
P3T-B-156
NEUTRONIC ANALYSIS OF ITER NEUTRAL BEAM TEST BED ------------------------------------------------------------------------ 45
P3T-B-160
A REVIEW OF JET NEUTRAL BEAM SYSTEM PERFORMANCE 1994 TO 2003 ----------------------------------------------------- 46
P3T-B-165
DEVELOPING A FULL SCALE ECRH MM-WAVE LAUNCHING SYSTEM MOCK-UP FOR ITER ------------------------------ 47
P3T-B-171
DIGITAL MOCK-UP DESIGN OF THE REMOTE STEERABLE ITER ECRH LAUNCHING SYSTEM --------------------------- 48
P3T-B-172
PRE STUDY RESULTS ON HIGH VOLTAGE SOLID-STATE SWITCHES FOR GYROTRON PROTECTION. ----------------- 49
P3T-B-173
AN ALTERNATIVE ECRH FRONT STEERING LAUNCHER FOR THE ITER UPPER PORT --------------------------------------- 50
P3T-B-174
DESIGN OF THE MM-WAVE SYSTEM OF THE ECRH UPPER LAUNCHER FOR ITER -------------------------------------------- 51
P3T-B-175
DEVELOPING THE NEXT LHCD SOURCE FOR TORE SUPRA -------------------------------------------------------------------------- 52
P3T-B-180
TOWARD AN LHCD SYSTEM FOR ITER ------------------------------------------------------------------------------------------------------ 53
P3T-B-181
A N-PORT ERROR MODEL AND CALIBRATION PROCEDURE FOR MEASURING THE SCATTERING MATRICES OF
LOWER-HYBRID MULTIJUNCTIONS ---------------------------------------------------------------------------------------------------------- 54
P3T-B-182
DESIGN AND FABRICATION OF THE "ITER-LIKE" SINGAP D¯ ACCELERATION SYSTEM ----------------------------------- 55
P3T-B-185
OPERATIONAL EXPERIENCE WITH UPGRADED JET NEUTRAL BEAM INJECTION SYSTEMS ----------------------------- 56
P3T-B-187
MAST NEUTRAL BEAM LONG PULSE UPGRADE------------------------------------------------------------------------------------------ 57
P3T-B-188
THE ITER NEUTRAL BEAM TEST FACILITY : DESIGNS OF THE GENERAL INFRASTRUCTURE, CRYOSYSTEM AND
COOLING PLANT ------------------------------------------------------------------------------------------------------------------------------------ 58
P3T-B-201
PROGRESS OF THE KSTAR ICRF COMPONENTS DEVELOPMENT FOR LONG PULSE OPERATION ------------------------ 59
P3T-B-204
RECENT PROGRESS OF NEGATIVE ION BASED NEUTRAL BEAM INJECTOR FOR JT-60U ----------------------------------- 60
P3T-B-210
TESTS AND FIRST RESULTS OF A LOAD RESILIENT ICRH ANTENNA ON TEXTOR ------------------------------------------- 61
P3T-B-211
REALISATION OF A TEST FACILITY FOR THE ICRH ITER PLUG-IN BY MEANS OF A MOCK-UP WITH SALTED
WATER LOAD ----------------------------------------------------------------------------------------------------------------------------------------- 62
P3T-B-218
STUDY OF MUTUAL COUPLING EFFECTS IN THE ANTENNA ARRAY OF THE ICRH PLUG-IN FOR ITER --------------- 63
P3T-B-221
O3A-B-221
STATUS AND PLANS FOR THE DEVELOPMENT OF AN RF NEGATIVE ION SOURCE FOR ITER NBI --------------- 64
P3T-B-225
DEVELOPMENT AND CONTRIBUTION OF RF HEATING AND CURRENT DRIVE SYSTEMS TO LONG PULSE,
HIGH PERFORMANCE EXPERIMENTS IN JT-60U -------------------------------------------------------------------------------------- 65
O3A-B-225
P3T-B-227
RF-SOURCE DEVELOPMENT FOR ITER: LARGE AREA H- BEAM EXTRACTION, MODIFICATIONS FOR LONG PULSE
OPERATION AND DESIGN OF A HALF SIZE ITER SOURCE ----------------------------------------------------------------------------- 66
P3T-B-229
DIAGNOSTICS AND MODELING OF THE PLASMA IN BATMAN RADIO FREQUENCY ION SOURCE ---------------------- 67
P3T-B-246
ECH MW-LEVEL CW TRANSMISSION LINE COMPONENTS SUITABLE FOR ITER ---------------------------------------------- 68
P3T-B-267
STATUS OF THE TJ-II ELECTRON BERNSTEIN WAVES HEATING PROJECT ------------------------------------------------------ 69
P3T-B-283
THE ASDEX UPGRADE ICRF SYSTEM: OPERATIONAL EXPERIENCE AND DEVELOPMENTS ------------------------------ 70
P3T-B-300
COOLING CONCEPTS OF THE ECRH LAUNCHER STRUCTURE AND THE TORUS WINDOWS ------------------------------ 71
P3T-B-310
THE DESIGN OF THE CONTROL SYSTEM FOR THE NEUTRAL BEAM INJECTION IN HT-7----------------------------------- 72
P3T-B-312
EXPERIMENTAL STUDY ON UNIFORMITY OF H- ION BEAM IN A LARGE NEGATIVE ION SOURCE --------------------- 73
P3T-B-314
DEVELOPMENT OF RELIABLE DIAMOND WINDOW FOR EC LAUNCHER ON FUSION REACTORS ----------------------- 74
P3T-B-320
DESIGN OF HIGH POWER COAXIAL DC BREAK FOR ADITYA TOKAMAK-------------------------------------------------------- 75
P3T-B-332
W7-X NEUTRAL-BEAM-INJECTION: TRANSMISSION, POWER-LOAD TO THE DUCT AND INNER VESSEL AND
CONSEQUENCES OF THE STELLARATOR STRAY FIELD ------------------------------------------------------------------------------- 76
4
P3T-B-337
DIAGNOSTICS OF THE CESIUM AMOUNT IN A RF NEGATIVE ION SOURCE AND THE CORRELATION WITH THE
EXTRACTED CURRENT DENSITY -------------------------------------------------------------------------------------------------------------- 77
P3T-B-345
LONG PULSE OPERATION ON THE KAMABOKO III ION SOURCE. ------------------------------------------------------------------ 78
P3T-B-351
DESIGN AND TEST OF A HV DEVICE FOR PROTECTION AND POWER MODULATION OF 140 GHZ/1MW CWGYROTRONS USED FOR ECRH ON W7-X ---------------------------------------------------------------------------------------------------- 79
P3T-B-353
DEVELOPMENT OF CW AND SHORT-PULSE CALORIMETRIC LOADS FOR HIGH POWER MILLIMETER-WAVE
BEAMS -------------------------------------------------------------------------------------------------------------------------------------------------- 80
P3T-B-356
ITER-LIKE PAM LAUNCHER FOR TORE SUPRA’S LHCD SYSTEM ------------------------------------------------------------------- 81
P3T-B-359
LARGE CRYOSORPTION PUMP OF THE TEST STAND FOR THE KSTAR NBI SYSTEM ----------------------------------------- 82
P3T-B-362
DEVELOPMENT OF A RF SOURCE FOR ITER NBI: FIRST RESULTS WITH D- OPERATION ----------------------------------- 83
P3T-B-364
PERFORMANCE TEST OF THE LH ANTENNA WITH CARBON GRILL IN JT-60U ------------------------------------------------- 84
P3T-B-371
FIRST RESULTS OF THE TORE SUPRA ITER LIKE ICRF ANTENNA PROTOTYPE ------------------------------------------------ 85
P3T-B-372
TORE SUPRA ITER-LIKE ANTENNA CHARACTERIZATION BY FEM ANALYSIS ------------------------------------------------ 86
P3T-B-382
MAINTENANCE SCHEMES FOR THE ITER NEUTRAL BEAM INJECTOR TEST FACILITY ------------------------------------- 87
P3T-B-385
NEUTRAL BEAM INJECTION OPTIMIZATION AT TJ-II ----------------------------------------------------------------------------------- 88
P3T-B-387
STATUS OF THE 140 GHZ / 10 MW CW TRANSMISSION SYSTEM FOR ECRH ON THE STELLARATOR W7-X ---------- 89
P3T-B-392
O3A-B-392
THE TEST OF A PAM LAUNCHER ON FTU: THE FIRST STEP TOWARD THE LHCD LAUNCHER FOR ITER ----- 90
P3T-B-410
MATERIAL PROCESSING AND PROTOTYPE FABRICATION OF HEAT TRANSFER ELEMENTS FOR SST-1 NBI
SYSTEM. ------------------------------------------------------------------------------------------------------------------------------------------------ 91
P3T-B-412
AN ALTERNATIVE SCHEME FOR THE ITER NBI POWER SUPPLY SYSTEM WITHDRAWN ---------------------------------- 92
P3T-B-439
NEUTRONICS ANALYSIS OF THE ECW LAUNCHING SYSTEM IN THE ITER UPPER PORT ----------------------------------- 93
P3T-B-455
THE ITER-LIKE ICRF ANTENNA FOR JET ---------------------------------------------------------------------------------------------------- 94
P3T-B-460
EFFECTS OF MUTUAL COUPLING ON ICRF LOAD-TOLERANT ANTENNAS ------------------------------------------------------ 95
P3T-B-479
140-GHZ HIGH-POWER GYROTRON DEVELOPMENT FOR THE STELLARATOR W7-X ---------------------------------------- 96
P3T-B-501
DEVELOPMENT OF THE 140 GHZ GYROTRON AND ITS SUBSYSTEMS FOR ECH AND ECCD IN TEXTOR -------------- 97
P3T-B-507
DESIGN OF CRYOSORPTION PUMPS FOR TESTBEDS OF ITER RELEVANT NEUTRAL BEAM INJECTORS -------------- 98
P3T-B-512
STATUS OF THE NEW ECRH SYSTEM FOR ASDEX UPGRADE ------------------------------------------------------------------------ 99
P3T-B-514
IMPROVED 118 GHZ GYROTRON FOR ECRH EXPERIMENTS ON TORE SUPRA ------------------------------------------------ 100
P3T-B-542
O3A-B-542
MATCHING TO ELMY PLASMAS IN THE ICRF DOMAIN --------------------------------------------------------------------------- 101
5
- C - Plasma Engineering and Control. ------------------------------------------------------------------------------------------------------------------------------ 102
P3C-C-62
OPTIMISED MODELLING OF THE TORE SUPRA TOKAMAK FOR PLASMA EQUILIBRIUM CALCULATIONS WITH
THE PROTEUS CODE ------------------------------------------------------------------------------------------------------------------------------ 102
P3C-C-77
HIGH PERFORMANCE INTEGRATED PLASMA CONTROL IN DIII–D* -------------------------------------------------------------- 103
P3C-C-79
PROGRESS TOWARDS ACHIEVING PROFILE CONTROL IN THE RECENTLY UPGRADED DIII-D PLASMA CONTROL
SYSTEM* ---------------------------------------------------------------------------------------------------------------------------------------------- 104
P3C-C-103
REAL TIME CONTROL OF FULLY NON-INDUCTIVE OPERATION IN TORE SUPRA LEADING TO 1GJ PLASMA
DISCHARGES---------------------------------------------------------------------------------------------------------------------------------------- 105
O1B-C-103
P3C-C-114
FEEDBACK CONTROL FOR PLASMA POSITION IN HL-2A TOKAMAK ------------------------------------------------------------- 106
P3C-C-149
DIII-D INTEGRATED PLASMA CONTROL TOOLS APPLIED TO NEXT GENERATION TOKAMAKS* ---------------------- 107
P3C-C-155
CONFIGURATION AND PERTURBATION DEPENDENCE OF THE NEUTRAL POINT IN JET --------------------------------- 108
P3C-C-157
DEVELOPMENT OF THE DINA-CH FULL DISCHARGE TOKAMAK SIMULATOR ----------------------------------------------- 109
P3C-C-161
DESIGN, IMPLEMENTATION AND TEST OF THE EXTREME SHAPE CONTROLLER (XSC) IN JET ------------------------- 110
P3C-C-190
CORRECTION POSSIBILITIES OF MAGNETIC FIELD ERRORS IN WENDELSTEIN 7-X ---------------------------------------- 111
P3C-C-207
REAL TIME CONTROL ENVIRONMENT IN THE RFX EXPERIMENT ---------------------------------------------------------------- 112
P3C-C-233
A FAST AND VERSATILE INTERLOCK SYSTEM ------------------------------------------------------------------------------------------ 113
P3C-C-268
NEW VISUALIZATION SYSTEM FOR CONTROLLING AND MONITORING PURPOSES IN THE TJ-II STELLARATOR 114
P3C-C-277
WEB-BASED GROUND LOOP SUPERVISION SYSTEM FOR THE TJ-II STELLARATOR ---------------------------------------- 115
P3C-C-291
A NEW CONTROLLER FOR THE JET ERROR FIELD CORRECTION COILS -------------------------------------------------------- 116
P3C-C-298
USING REAL TIME WORKSHOP FOR RAPID AND RELIABLE CONTROL IMPLEMENTATION IN THE FRASCATI
TOKAMAK UPGRADE FEEDBACK CONTROL SYSTEM RUNNING UNDER RTAI-LINUX ------------------------------------ 117
P3C-C-301
THE SYSTEM ARCHITECTURE OF THE NEW JET SHAPE CONTROLLER --------------------------------------------------------- 118
P3C-C-350
AN INTEGRAL APPROACH TO PLASMA SHAPE CONTROL --------------------------------------------------------------------------- 119
P3C-C-352
LINEARIZED MODELS OF THE PLASMA RESPONSE IN THE NEW RFX LOAD ASSEMBLY --------------------------------- 120
P3C-C-354
DESIGN OF THE NEW RFX EQUILIBRIUM ACTIVE CONTROL SYSTEM ---------------------------------------------------------- 121
P3C-C-363
O1B-C 363
REAL-TIME MEASUREMENT AND CONTROL AT JET- EXPERIMENT CONTROL ---------------------------------------- 122
P3C-C-375
OPEN LOOP CHARACTERIZATION OF AN ACTIVE CONTROL SYSTEM OF MHD MODES ---------------------------------- 123
P3C-C-377
COMPARISON OF STRATEGIES AND REGULATOR DESIGN FOR ACTIVE CONTROL OF MHD MODES ---------------- 124
P3C-C-383
ADOPTING MODERN NONLINEAR CONTROL TECHNIQUES FOR THE PLASMA STABILIZATION ON THE NOVEL
LINUX-BASED FEEDBACK CONTROLLER OF FTU --------------------------------------------------------------------------------------- 125
P3C-C-403
VERTICAL STABILITY OF ITER PLASMAS WITH 3D PASSIVE STRUCTURES AND A DOUBLE LOOP CONTROL
SYSTEM ------------------------------------------------------------------------------------------------------------------------------------------------ 126
P3C-C-408
THE BASIC METHODS FOR UNDERSTANDING OF PLASMA EQUILIBRIUM TOWARD ADVANCED CONTROL ------ 127
P3C-C-457
XSC PLASMA CONTROL: TOOL DEVELOPMENT FOR THE SESSION LEADER -------------------------------------------------- 128
P3C-C-463
A FLEXIBLE AND REUSABLE SOFTWARE FOR REAL-TIME CONTROL APPLICATIONS AT JET -------------------------- 129
P3C-C-508
COMMISSIONING TESTS FOR CONTROL PROCESSES IN ASDEX UPGRADE´S NEW CONTROL AND DATA
ACQUISITION SYSTEM --------------------------------------------------------------------------------------------------------------------------- 130
P3C-C-510
OPTIMIZATION OF THE IGNITOR OPERATING SCENARIO AT 11 MA ------------------------------------------------------------- 131
P3C-C-517
PLASMA FEEDBACK CONTROLLER REORGANISATION FOR ASDEX UPGRADE'S NEW DISCHARGE CONTROL AND
DATA ACQUISITION SYSTEM ------------------------------------------------------------------------------------------------------------------ 132
6
- D - Diagnostics, Data Acquisition and Remote Participation. ----------------------------------------------------------------------------------------------- 133
P2C-D-22
NEUTRON ANALYSIS OF H-ALPHA AND CXRS DIAGNOSTICS OF ITER ---------------------------------------------------------- 133
P2C-D-42
NEW CONSTRAINTS FOR PLASMA DIAGNOSTICS DEVELOPMENT DUE TO THE HARSH ENVIRONMENT OF MJ
CLASS LASERS -------------------------------------------------------------------------------------------------------------------------------------- 134
P2C-D-52
THE INTEGRATED VISUALISATION SOFTWARE FOR THE ITER IN VESSEL VIEWING SYSTEM (IVVS) --------------- 135
P2C-D-60
USING REMOTE PARTICIPATION TOOLS TO IMPROVE COLLABORATIONS ---------------------------------------------------- 136
P2C-D-74
CALORIMETRY MEASUREMENTS DURING HIGH ENERGY DISCHARGES AT TORE SUPRA------------------------------- 137
P2C-D-80
REAL-TIME MULTIPLE NETWORKED VIEWER CAPABILITY OF THE DIII-D EC DATA ACQUISITION SYSTEM*---- 138
P2C-D-91
EXPERIMENTAL STUDY OF RADIATION-INDUCED CURRENTS IN COPPER AND STAINLESS STEEL CORE
MINERAL-INSULATED CABLES IN THE BR2 RESEARCH REACTOR --------------------------------------------------------------- 139
P2C-D-98
TORE-SUPRA INFRARED THERMOGRAPHY SYSTEM, A REAL STEADY STATE DIAGNOSTIC. --------------------------- 140
P2C-D-104
NEW INSTRUMENTS FOR ADVANCED NEUTRON EMISSION SPECTROMETRY DIAGNOSIS OF D AND DT PLASMAS
AT JET -------------------------------------------------------------------------------------------------------------------------------------------------- 141
P2C-D-119
SURFACE DIAGNOSTICS WITH APPLICATION OF VIDEOSCOPE ON THE BASIS OF CU-LASER -------------------------- 142
P2C-D-120
NEW MANAGING SYSTEM OF A LARGE AMOUNT OF IMAGES ON TORE SUPRA--------------------------------------------- 143
P2C-D-132
RADIATION RESISTANT BOLOMETERS USING PLATINUM ON AL2O3 AND ALN --------------------------------------------- 144
P2C-D-142
TEMPERATURE DEPENDENCE OF THE TRANSMISSION LOSS IN KU-1 AND KS-4V QUARTZ GLASSES FOR THE
ITER DIAGNOSTIC WINDOW-------------------------------------------------------------------------------------------------------------------- 145
P2C-D-159
THERMAL AND NEUTRON TESTS OF MULTILAYERED DIELECTRIC MIRRORS ----------------------------------------------- 146
P2C-D-166
LASER DAMAGE INVESTIGATIONS OF CU MIRRORS ---------------------------------------------------------------------------------- 147
P2C-D-194
DESIGN OF LOST ALPHA PARTICLE DIAGNOSTICS FOR JET* ----------------------------------------------------------------------- 148
P2C-D-202
DEVELOPMENT OF THE PHASE COUNTER WITH THE REAL-TIME FRINGE JUMP CORRECTOR FOR
INTERFEROMETER ON LHD --------------------------------------------------------------------------------------------------------------------- 149
P2C-D-208
DATA ACQUISITION UPGRADE IN THE RFX EXPERIMENT --------------------------------------------------------------------------- 150
P2C-D-220
THERMAL DETECTOR FOR THE LOST ALPHA PARTICLE MEASUREMENTS --------------------------------------------------- 151
P2C-D-226
ITER RELEVANT DEVELOPMENTS OF NEUTRON DIAGNOSTICS DURING JET TRACE TRITIUM CAMPAIGN ------- 152
P2C-D-230
OPTICAL AND ELECTRICAL DEGRADATION OF HYDROGEN IMPLANTED KS-4V QUARTZ GLASS -------------------- 153
P2C-D-234
RECONSTRUCTION CAPABILITY OF JET MAGNETIC SENSORS -------------------------------------------------------------------- 154
P2C-D-235
QUENCH DETECTION & DATA ACQUISITION SYSTEM FOR SST-1 SUPERCONDUCTING MAGNETS ------------------- 155
P2C-D-248
LASER DAMAGE OF KU-1 SILICA GLASS COVERED WITH HYDROCARBON FILM ------------------------------------------- 156
P2C-D-250
WIDE AREA DATA REPLICATION IN AN ITER-RELEVANT DATA ENVIRONMENT ------------------------------------------- 157
P2C-D-251
O2B-D-251
ADVANCES IN REMOTE PARTICIPATION FOR FUSION EXPERIMENTS* --------------------------------------------------- 158
P2C-D-258
SOFT COMPUTING AND CHAOS TEORY FOR ANTICIPATION OF DISRUPTION IN TOKAMAK REACTORS ----------- 159
P2C-D-263
ADSORPTION IN INSULATOR MATERIALS ENHANCED BY D IMPLANTATION ----------------------------------------------- 160
P2C-D-270
RADIATION ENHANCED DEGRADATION OF SIO OVERCOATED ALUMINIUM MIRRORS ---------------------------------- 161
P2C-D-272
THE NEW MEASUREMENT MONITORING SYSTEM ON FTU -------------------------------------------------------------------------- 162
P2C-D-281
FIRST RESULTS OF MINIMUM FISHER REGULARISATION AS UNFOLDING METHOD FOR JET NE213 LIQUID
SCINTILLATOR NEUTRON SPECTROMETRY ---------------------------------------------------------------------------------------------- 163
P2C-D-282
PRESENT STATUS OF THE TJ-II REMOTE PARTICIPATION SYSTEM--------------------------------------------------------------- 164
P2C-D-347
APD DETECTOR ELECTRONICS AND PXI BASED DATA ACQUISITION SYSTEM FOR SST-1 THOMSON SCATTERING
DIAGNOSTICS --------------------------------------------------------------------------------------------------------------------------------------- 165
P2C-D-366
REAL TIME MEASUREMENT AND CONTROL AT JET - DIAGNOSTIC SYSTEMS ----------------------------------------------- 166
P2C-D-370
O2B-D-370
OPTICAL FIBERS FOR PLASMA DIAGNOSTICS UNDER GAMMA-RAY AND UV IRRADIATION --------------------- 167
P2C-D-384
THE HALO CURRENT SENSOR SYSTEM FOR JET-EP ------------------------------------------------------------------------------------ 168
P2C-D-396
DEVELOPMENT OF ACTIVELY COOLED PERISCOPES FOR DIVERTOR OBSERVATION ------------------------------------ 169
P2C-D-398
DIAGNOSTICS FOR STUDYING DEPOSITION AND EROSION PROCESSES IN JET ---------------------------------------------- 170
P2C-D-428
VULNERABILITY OF OPTICAL FIBERS FOR PLASMA DIAGNOSTICS OF LASER MEGAJOULE --------------------------- 171
7
P2C-D-433
APPLICATION OF ORTHOGONALLY POLARIZED TWO-FREQUENCY LASER TO POLARIMETER FOR MAGNETIC
FIELD MEASUREMENTS OF LONG-PULSED FUSION DEVICES WITHDRAWN -------------------------------------------------- 172
P2C-D-437
THE TEXTOR DIAGNOSTIC DATA MANAGEMENT CHAIN ---------------------------------------------------------------------------- 173
P2C-D-446
NEW LOW LOSS TRIAXIAL AND MAGNETICS DIAGNOSTICS FEEDTHROUGH AT JET-------------------------------------- 174
P2C-D-451
THERMO-STRESS ANALYSIS OF OPTICAL MATERIALS FOR HIGH HEAT FLUX APPLICATIONS ------------------------ 175
P2C-D-458
DESIGN AND MANUFACTURE OF THE UPPER COILS AND OUTER POLOIDAL COILS SUBSYSTEMS FOR THE JET-EP
MAGNETIC DIAGNOSTIC ------------------------------------------------------------------------------------------------------------------------ 176
P2C-D-459
DESIGN OF EX-VESSEL MAGNETIC PROBES FOR JET-EP ----------------------------------------------------------------------------- 177
P2C-D-466
TRANSDUCERS AND SIGNAL CONDITIONERS OF THE RFX NEW MAGNETIC MEASUREMENT SYSTEM ------------ 178
P2C-D-476
WIDE-ANGLE INFRARED THERMOGRAPHY FOR JET-EP ------------------------------------------------------------------------------ 179
P2C-D-477
LITHIUM BEAM DEVELOPMENTS FOR HIGH-ENERGY PLASMA DIAGNOSTICS ---------------------------------------------- 180
P2C-D-502
THE NEW TAE - ALFVÉN WAVE ACTIVE EXCITATION SYSTEM AT JET --------------------------------------------------------- 181
P2C-D-503
NEW MILLIMETER-WAVE ACCESS FOR JET REFLECTOMETRY AND ECE ------------------------------------------------------ 182
P2C-D-504
CONTROL PROCESS STRUCTURE OF ASDEX UPGRADE´S NEW CONTROL AND DATA ACQUISITION SYSTEM
------------------------------------------------------------------------------------------------------------------------------------------------------------ 183
O2B-D-504
P2C-D-506
MULTI-SUPPORT VECTOR MACHINES FOR DISRUPTION CLASSIFICATION IN TOKAMAK REACTORS --------------- 184
P2C-D-513
OPTICAL DESIGN OF THE OBLIQUE ECE ANTENNA SYSTEM FOR JET ----------------------------------------------------------- 185
P2C-D-515
ITER DIAGNOSTICS: MAINTENANCE AND COMMISSIONING IN THE HOT CELL TEST BED------------------------------- 186
P2C-D-519
NEW BOLOMETRY CAMERAS FOR THE JET ENHANCED PERFORMANCE PHASE -------------------------------------------- 187
8
- E - Magnets and Power Supplies. ---------------------------------------------------------------------------------------------------------------------------------- 188
P2T-E-9
THE BATCH PRODUCTION FOR SUPERCONDUCTING MAGNET COILS OF EAST (HT-7U) ---------------------------------- 188
P2T-E-20
STUDY ON HIGH-POWER HIGH-FREQUENCY INVERTER FOR FAST PLASMA POSITION CONTROL IN EAST SUPERCONDUCTING TOKAMAK ----------------------------------------------------------------------------------------------------------------------- 189
P2T-E-23
A LOW COST JOINT FOR THE ITER PF COILS, DESIGN AND TEST RESULTS. --------------------------------------------------- 190
P2T-E-30
130KV 130A HIGH VOLTAGE SWITCHING MODE POWER SUPPLY FOR NEUTRAL INJECTORS - CONTROL ISSUES
AND ALGORITHMS -------------------------------------------------------------------------------------------------------------------------------- 191
P2T-E-31
FIBERGLASS UNIDIRECTIONAL COMPOSITE TO BE USED FOR ITER PRE-COMPRESSION RINGS ---------------------- 192
P2T-E-34
MEASUREMENT OF CONTACT RESISTANCE DISTRIBUTION IN TYPICAL ITER SIZE CONDUCTOR TERMINATION
------------------------------------------------------------------------------------------------------------------------------------------------------------ 193
P2T-E-35
UPDATING THE DESIGN OF THE FEEDER COMPONENTS FOR THE ITER MAGNET SYSTEM ------------------------------ 194
P2T-E-36
MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: COMPUTATION
OF THE BACKGROUND FIELD AND CONSEQUENCES ON THE DESIGN OF THE ELECTRICAL DISTRIBUTION
BOARDS AND CONTROL BOARDS FOR THE ITER TOKAMAK BUILDING ------------------------------------------------------- 195
P2T-E-37
COMMISSIONING OF THE 10 POWER SUPPLIES OF THE CONTROL COILS OF WENDELSTEIN 7-X EXPERIMENT -- 196
P2T-E-40
DESIGN AND COMMISSIONING OF THE NEW TOROIDAL FIELD COIL FOR THE NATIONAL SPHERICAL TORUS
EXPERIMENT (NSTX) ------------------------------------------------------------------------------------------------------------------------------ 197
P2T-E-48
ANALYSES AND IMPLICATIONSOF V-I CHARACTERISTIC --------------------------------------------------------------------------- 198
P2T-E-50
PIONEERING SUPERCONDUCTING MAGNETS IN LARGE TOKAMAKS: EVALUATION AFTER 17 YEARS OF
OPERATING EXPERIENCE --------------------------------------------------------------------------------------------------------------------- 199
O2A-E-50
P2T-E-55
STABILITY, THERMAL EQUILIBRIUM AND DESIGN CRITERIA FOR CABLE-IN-CONDUIT-CONDUCTORS WITH A
BROAD TRANSITION TO NORMAL STATE ------------------------------------------------------------------------------------------------- 200
P2T-E-68
DESIGN OPTIMISATION OF THE ITER TF COIL CASE AND STRUCTURES ------------------------------------------------------- 201
P2T-E-73
FABRICATION OF THE PLANAR COILS FOR WENDELSTEIN 7-X ------------------------------------------------------------------- 202
P2T-E-81
OVERVIEW OF THE DIII–D INTERNAL RESISTIVE WALL MODE STABILIZATION POWER SUPPLY SYSTEM* ------ 203
P2T-E-95
THE EUROPEAN DEVELOPMENT OF A FULL SCALE SWITCHING UNIT FOR THE ITER SWITCHING AND
DISCHARGING NETWORKS ------------------------------------------------------------------------------------------------------------------- 204
O2A-E-95
P2T-E-97
MECHANICAL PERFORMANCE OF MAGNET INSULATION MATERIALS FABRICATED BY THE “INSULATE-WINDAND-REACT “ TECHNIQUE* -------------------------------------------------------------------------------------------------------------------- 205
P2T-E-99
INFLUENCE OF PARAMETER VARIATIONS ON THE FATIGUE BEHAVIOR OF MAGNET INSULATION SYSTEMS -- 206
P2T-E-101
MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: TESTS ON
PROGRAMMABLE LOGICAL CONTROLLERS AND OTHER ELECTRONIC DEVICES ------------------------------------------ 207
P2T-E-105
DESIGN, FABRICATION AND INSTALLATION OF CRYOGENIC TARGET SYSTEM FOR 14 MEV NEUTRON
IRRADIATION ---------------------------------------------------------------------------------------------------------------------------------------- 208
P2T-E-106
THE EUROPEAN NB3SN ADVANCED STRAND DEVELOPMENT PROGRAMME ------------------------------------------------ 209
P2T-E-111
DESIGN AND DEVELOPMENT OF THE POWER SUPPLY SYSTEM FOR HL-2A TOKAMAK ---------------------------------- 210
P2T-E-112
THE ITER THERMAL SHIELDS FOR THE MAGNET SYSTEM: DESIGN EVOLUTION AND ANALYSIS -------------------- 211
P2T-E-121
QUALITY ASSURANCE PROCEDURES IN THE EAST MAGNETS MANUFACTURING PROCESS --------------------------- 212
P2T-E-126
THYRISTOR CROWBAR SYSTEM FOR THE HIGH CURRENT POWER SUPPLIES OF ASDEX UPGRADE ----------------- 213
P2T-E-186
OPTIMIZATION OF THE POWER SUPPLY FOR A HELIAS REACTOR SUPERCONDUCTING COIL SYSTEM ------------ 214
P2T-E-198
QUENCH CURRENT MEASUREMENT AND PERFORMANCE EVALUATION OF THE EAST TOROIDAL FIELD
COILS -------------------------------------------------------------------------------------------------------------------------------------------------- 215
O2A-E-198
P2T-E-209
THERMAL AND STRUCTURAL ANALYSIS OF THE W7-X MAGNET HEAT RADIATION SHIELD -------------------------- 216
P2T-E-212
FILAMENT POWER SUPPLY (AC TO AC CONVERTER) FOR LONG PULSE NEUTRAL BEAM INJECTOR OF SST-1 --- 217
P2T-E-219
TRANSIENT ELECTRICAL BEHAVIOUR OF THE ITER TF COILS DURING FAST DISCHARGE AND TWO FAULT
CASES -------------------------------------------------------------------------------------------------------------------------------------------------- 218
P2T-E-223
STUDIES ON THE BEHAVIOR OF MULTISECONDARY TRANSFORMERS USED FOR REGULATED HV POWER
SUPPLIES ---------------------------------------------------------------------------------------------------------------------------------------------- 219
P2T-E-236
HIGH POWER IGBT BRIGE APPLICATION FOR THE HARMONIC SUPPRESSION IN THE POWER SUPPLY SYSTEM OF
THE SPANISH STELLARATOR TJ-II. ---------------------------------------------------------------------------------------------------------- 220
P2T-E-259
MANUFACTURE AND TEST OF THE NON-PLANAR COILS FOR WENDELSTEIN 7-X ------------------------------------------ 221
9
P2T-E-266
V-I CHARACTERISTICS WITH BUMPS IN THE MEDIUM SIZE NBTI CICC CABLES. ------------------------------------------- 222
P2T-E-284
O2A-E-284
HIGH TEMPERATURE SUPERCONDUCTORS FOR THE ITER MAGNET SYSTEM AND BEYOND -------------------- 223
P2T-E-299
ANALYSIS OF THE RESISTIVE TRANSITION IN NB-TI CABLE-IN-CONDUIT CONDUCTORS VIA AN EXTENDED 1-D
MODEL ------------------------------------------------------------------------------------------------------------------------------------------------- 224
P2T-E-304
EFFECTIVE BENDING STRAIN ESTIMATED FROM IC TEST RESULT OF D SHAPED NB3AL CICC COIL FABRICATED
WITH A REACT AND WIND PROCESS FOR THE NATIONAL CENTRALIZED TOKAMAK ------------------------------------ 225
P2T-E-306
ELIMINATION OF VARIABLE HARMONICS ON MOTOR GENERATOR CIRCUIT FOR EXPERIMENTAL FUSION
FACILITY ---------------------------------------------------------------------------------------------------------------------------------------------- 226
P2T-E-311
FATIGUE ASSESSMENT OF THE ITER TF COIL CASE BASED ON JJ1 FATIGUE TESTS --------------------------------------- 227
P2T-E-326
EFFECT OF ELECTRICAL CHARACTERISTICS OF SIC POWER DEVICE ON OPERATIONAL EFFICIENCY OF AC/DC
CONVERTER ----------------------------------------------------------------------------------------------------------------------------------------- 228
P2T-E-338
DESIGN REQUIREMENT, QUALIFICATION TESTS AND INTEGRATION OF A THIN SOLID LUBRICANT FILM OF
MOS2 FOR COLD MASS SUPPORT STRUCTURE OF THE STEADY STATE SUPERCONDUCTING TOKAMAK SST-1. 229
P2T-E-344
HOW SHOULD WE TEST THE ITER TF COILS ? -------------------------------------------------------------------------------------------- 230
P2T-E-379
CYCLIC TESTING OF SHEAR KEYS FOR THE ITER MAGNET SYSTEM ------------------------------------------------------------ 231
P2T-E-390
MODULAR COIL DESIGN DEVELOPMENTS FOR THE NATIONAL COMPACT STELLARATOR EXPERIMENT (NCSX)
------------------------------------------------------------------------------------------------------------------------------------------------------------ 232
P2T-E-394
CONCEPTUAL DESIGN OF SPHERICAL TORUS WITH TF-CS HYBRID COILS BASED ON VIRIAL THEOREM --------- 233
P2T-E-405
EMI ON DIAGNOSTICS AND CONTROL CIRCUITS DUE TO SWITCHING POWER SUPPLIES -------------------------------- 234
P2T-E-406
THE CONTROL SYSTEM OF THE TOROIDAL POWER SUPPLY OF RFX ------------------------------------------------------------ 235
P2T-E-417
COMPONENTS AND SYSTEM TESTS ON THE RFX TOROIDAL POWER SUPPLY ----------------------------------------------- 236
P2T-E-420
COMMISSIONING AND OPERATION OF 130KV/130A SWITCHED-MODE HV POWER SUPPLIES WITH THE
UPGRADED JET NEUTRAL BEAM INJECTORS -------------------------------------------------------------------------------------------- 237
P2T-E-427
MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR ELECTRICAL INSTALLATIONS: TESTS ON LOW
VOLTAGE CIRCUIT BREAKERS AND CONTACTORS------------------------------------------------------------------------------------ 238
P2T-E-442
FIRST INTEGRATED TEST OF THE SUPERCONDUCTING MAGNET SYSTEMS FOR THE LEVITATED DIPOLE
EXPERIMENT (LDX) ------------------------------------------------------------------------------------------------------------------------------- 239
P2T-E-462
MODELING AC LOSSES IN THE ITER NBTI FULL SIZE JOINT SAMPLES USING THE THELMA CODE ------------------- 240
P2T-E-471
POWER DISSIPATION AND ENERGY TRANSFER DURING TESTING OF THE ITER TOROIDAL FIELD MODEL COIL241
P2T-E-490
DC AND TRANSIENT CURRENT DISTRIBUTION ANALYSIS FROM SELF-FIELD MEASUREMENTS ON ITER PFIS
CONDUCTOR ----------------------------------------------------------------------------------------------------------------------------------------- 242
P2T-E-491
THE MAGNET SYSTEM OF THE KTM TOKAMAK ---------------------------------------------------------------------------------------- 243
P2T-E-511
OPTIMISATION OF THE CURRENT DISTRIBUTION IN THE IGNITOR POLOIDAL FIELD COILS AND EVALUATION OF
THE COILS TEMPERATURES AND RESISTANCE DURING THE REFERENCE OPERATING SCENARIO ------------------ 244
P2T-E-528
SAFETY ASSESSMENT OF THE ITER COILS SYSTEM ----------------------------------------------------------------------------------- 245
P2T-E-534
WINDING MACHINES FOR THE MANUFACTURING OF SUPERCONDUCTIVE COILS OF THE MAIN EUROPEAN
FUSION RESEARCH MACHINES --------------------------------------------------------------------------------------------------------------- 246
P2T-E-539
A SUCCESS STORY: LHC CABLE PRODUCTION AT ALSTOM MSA ----------------------------------------------------------------- 247
10
- F - Plasma Facing Components. ------------------------------------------------------------------------------------------------------------------------------------ 248
P4C-F-8
TILES CHAMFERING AND POWER HANDLING OF THE MK II HD DIVERTOR -------------------------------------------------- 248
P4C-F-12
THERMAL AND MECHANICAL ANALYSIS OF THE EAST PLASMA FACING COMPONENTS ------------------------------- 249
P4C-F-13
THE DYNAMIC ERGODIC DIVERTOR IN TEXTOR – A NOVEL TOOL FOR STUDYING MAGNETIC PERTURBATION
FIELD EFFECTS -------------------------------------------------------------------------------------------------------------------------------------- 250
P4C-F-24
EAST(HT-7U) IN-VESSEL COMPONENTS DESIGN ---------------------------------------------------------------------------------------- 251
P4C-F-27
HOT RADIAL PRESSING: AN ALTERNATIVE TECHNIQUE FOR THE MANUFACTURING OF PLASMA-FACING
COMPONENTS --------------------------------------------------------------------------------------------------------------------------------------- 252
P4C-F-28
HETS PERFORMANCES IN HE COOLED POWER PLANT DIVERTOR --------------------------------------------------------------- 253
P4C-F-32
THE INFLUENCE OF IRRADIATION REGIMES ON RETENTION HYDROGEN ISOTOPES IN STRUCTURAL MATERIALS
------------------------------------------------------------------------------------------------------------------------------------------------------------ 254
P4C-F-33
MANUFACTURING TECHNOLOGY DEVELOPMENT FOR THE VACUUM VESSEL AND PLASMAFACING
COMPONENTS --------------------------------------------------------------------------------------------------------------------------------------- 255
P4C-F-41
ENGINEERING AND THERMAL-HYDRAULIC DESIGN OF PFC COOLING FOR SST-1 TOKAMAK ------------------------- 256
P4C-F-49
THE USE OF COPPER ALLOY CUCRZR AS A STRUCTURAL MATERIAL FOR ACTIVELY COOLED PLASMA FACING
AND IN VESSEL COMPONENTS ---------------------------------------------------------------------------------------------------------------- 257
P4C-F-54
MANUFACTURING OF THE W7-X DIVERTOR AND WALL PROTECTION --------------------------------------------------------- 258
P4C-F-59
STUDIES ON GRAPHITE SURFACES DETRITIATION BY PULSED REPETITION RATE NANOSECOND LASERS ------ 259
P4C-F-66
STEADY STATE AND TRANSIENT THERMAL-HYDRAULIC ANALYSES ON ITER DIVERTOR MODULE --------------- 260
P4C-F-69
APPLIED TECHNOLOGIES AND INSPECTIONS FOR THE W7-X PRE-SERIES TARGET ELEMENTS------------------------ 261
P4C-F-76
OVERVIEW OF THE ENGINEERING DESIGN OF THE ITER DIVERTOR ------------------------------------------------------------ 262
P4C-F-92
TOWARDS THE DEVELOPMENT OF WORKABLE ACCEPTANCE CRITERIA FOR THE DIVERTOR CFC MONOBLOCK
ARMOUR.---------------------------------------------------------------------------------------------------------------------------------------------- 263
P4C-F-100
RESULTS AND ANALYSIS OF HIGH HEAT FLUX TESTS ON A FULL SCALE VERTICAL TARGET PROTOTYPE OF
ITER DIVERTOR ------------------------------------------------------------------------------------------------------------------------------------- 264
P4C-F-115
STRUCTURAL AND FRACTURE MECHANICS ANALYSIS OF ITER TOROIDAL FIELD COIL -------------------------------- 265
P4C-F-116
CRACK PROPAGATION BEHAVIOR AROUND DSCU/SS316 HIP BONDED INTERFACE BY THERMAL FATIGUE ----- 266
P4C-F-133
SIMULATION OF MANY-ATOMIC INTERACTIONS IN W-O-H SYSTEM WITH THE MD CODE CADAC ------------------ 267
P4C-F-135
DESIGN OF A LIMITER FOR THE JET EP ICRH ANTENNA ----------------------------------------------------------------------------- 268
P4C-F-145
EROSION OF TUNGSTEN MACROBRUSH ARMOR AFTER MULTIPLE INTENSE TRANSIENT EVENTS IN ITER ------ 269
P4C-F-146
DEVELOPMENT OF AN ORIGINAL ACTIVE THERMOGRAPHY METHOD ADAPTED TO ITER PLASMA FACING
COMPONENTS CONTROL ------------------------------------------------------------------------------------------------------------------------ 270
P4C-F-162
PLASMA SPRAYED TUNGSTEN-BASED COATINGS AND THEIR PERFORMANCE UNDER FUSION RELEVANT
CONDITIONS ----------------------------------------------------------------------------------------------------------------------------------------- 271
P4C-F-167
HIGH TEMPERATURE STRESSES IN ITER RELEVANT BRAZED GLIDCOP/W MODEL STRUCTURES -------------------- 272
P4C-F-176
A MATURE INDUSTRIAL SOLUTION FOR ITER DIVERTOR PLASMA FACING COMPONENTS: HYPERVAPOTRON
COOLING CONCEPT ADAPTED TO TORE SUPRA FLAT TILE TECHNOLOGY --------------------------------------------------- 273
P4C-F-192
CONCEPTUAL DESIGN OD A HIGH-TEMPERATURE WATER-COOLED DIVERTOR FOR A FUSION POWER REACTOR
------------------------------------------------------------------------------------------------------------------------------------------------------------ 274
P4C-F-195
DEVELOPMENT OF A COPPER ALLOY TO BERYLLIUM HIP BONDING TECHNOLOGY FOR THE ITER FIRST WALL
------------------------------------------------------------------------------------------------------------------------------------------------------------ 275
P4C-F-228
AN ADVANCED HE-COOLED DIVERTOR CONCEPT: DESIGN, COOLING TECHNOLOGY, AND THERMOHYDRAULIC
ANALYSES WITH CFD ---------------------------------------------------------------------------------------------------------------------------- 276
P4C-F-239
THE NEW ELECTRON BEAM TEST FACILITY JUDITH II FOR HIGH HEAT FLUX EXPERIMENTS ON PLASMA FACING
COMPONENTS. -------------------------------------------------------------------------------------------------------------------------------------- 277
P4C-F-253
FORMATION OF CRYSTALLINE NANOSTRUCTURES DURING DEUTERIUM PLASMA INTERACTION WITH
TUNGSTEN-BASED MATERIALS IN SIMULATED GAS DIVERTOR CONDITIONS. --------------------------------------------- 278
P4C-F-265
ACTIVITY OF THE EUROPEAN HIGH HEAT FLUX TEST FACILITY: FE200 ------------------------------------------------------- 279
P4C-F-274
PROPOSAL OF LITIZATION OF FTU VACUUM VESSEL BY USING A LITHIUM LIMITER ------------------------------------ 280
P4C-F-278
DESIGN, PERFORMANCE AND CONSTRUCTION OF A 2 MW ION BEAM TEST FACILITY FOR PLASMA FACING
COMPONENTS --------------------------------------------------------------------------------------------------------------------------------------- 281
P4C-F-279
SPECTROSCOPIC STUDIES OF HOMOGENEOUS CARBON FLAKES WITH A HIGH DEUTERIUM CONTENT FORMED
IN TOKAMAK T-10---------------------------------------------------------------------------------------------------------------------------------- 282
11
P4C-F-280
VACUUM PLASMA-SPRAYED TUNGSTEN ON EUROFER AND 316L - RESULTS OF CHARACTERISATION AND
THERMAL LOADING TESTS - ------------------------------------------------------------------------------------------------------------------- 283
P4C-F-285
CAN TOKAMAK DEVICES SURVIVE ELMS DURING NORMAL OPERATION? A SIMULATION STUDY------------------ 284
P4C-F-294
O4A-F-294
EU R&D ON DIVERTOR COMPONENTS --------------------------------------------------------------------------------------------------- 285
P4C-F-305
THERMAL MODELING OF W ROD ARMOR SUBJECTED TO ELMS ------------------------------------------------------------------ 286
P4C-F-315
CRITICAL HEAT FLUX TESTING ON SCREW COOLING TUBE MADE OF RAFM-STEEL F82H FOR DIVERTOR
APPLICATION ---------------------------------------------------------------------------------------------------------------------------------------- 287
P4C-F-328
STATUS OF HE-COOLED DIVERTOR DEVELOPMENT FOR DEMO ------------------------------------------------------------------ 288
P4C-F-333
NUMERICAL AND EXPERIMENTAL STUDY OF DEMO HE-COOLED DIVERTOR TARGET MOCK-UPS ------------------ 289
P4C-F-343
MEASUREMENTS OF H/D DIFFUSIVITY IN AND SOLUBILITY THROUGH TUNGSTEN IN THE TEMPERATURE
RANGE OF 600 C TO 800 C------------------------------------------------------------------------------------------------------------------------ 290
P4C-F-367
TESTING OF ACTIVELY COOLED MOCK-UPS IN SEVERAL HIGH HEAT FLUX FACILITIES – AN INTERNATIONAL
ROUND ROBIN TEST ------------------------------------------------------------------------------------------------------------------------------- 291
P4C-F-376
STUDY OF TECHNOLOGICAL AND MATERIAL ASPECTS OF HE-COOLED DIVERTOR FOR DEMO REACTOR ------- 292
P4C-F-386
DESIGN AND THERMAL PERFORMANCE OF SURFACE-BOLTLESS MECHANICALLY ATTACHED MODULE FOR
DIVERTOR PLATE OF LHD ---------------------------------------------------------------------------------------------------------------------- 293
P4C-F-413
O4A-F-413
MANUFACTURE OF BLANKET SHIELD MODULES FOR ITER ------------------------------------------------------------------- 294
P4C-F-426
THE MAST IMPROVED DIVERTOR ------------------------------------------------------------------------------------------------------------ 295
P4C-F-429
OXYGEN IMPURITY EFFECTS ON HYDROGEN ISOTOPE RELEASE FROM PLASMA CHEMICAL VAPOR DEPOSITION
BORON COATING ----------------------------------------------------------------------------------------------------------------------------------- 296
P4C-F-431
IMPLANTATION TEMPERATURE DEPENDENCE ON DEUTERIUM BEHAVIOR IN HIGHLY ORIENTED PYROLITIC
GRAPHITE --------------------------------------------------------------------------------------------------------------------------------------------- 297
P4C-F-445
MANUFACTURING AND TESTING IN REACTOR RELEVANT CONDITIONS OF BRAZED PLASMA FACING
COMPONENTS OF THE ITER DIVERTOR------------------------------------------------------------------------------------------------- 298
O4A-F-445
P4C-F-447
DEVELOPMENT OF THE PLASMA FACING COMPONENTS FOR THE DOME-LINER COMPONENT OF THE ITER
DIVERTOR -------------------------------------------------------------------------------------------------------------------------------------------- 299
P4C-F-474
THERMAL PROPERTY CHANGES OF ERODED AND REPETITIVELY LOADED CFC ------------------------------------------- 300
P4C-F-475
STUDIES OF HEAT CONDUCTION IN LIQUID LITHIUM CAPILLARY POROUS SYSTEM ------------------------------------- 301
12
- G - Vessel-in vessel Engineering and Remote Handling. ---------------------------------------------------------------------------------------------------- 302
P4T-G-4
DESIGN AND DEVELOPMENT TOWARDS A PARALLEL WATER HYDRAULIC WELD/CUT ROBOT FOR
MACHINING PROCESSES IN ITER VACUUM VESSEL ------------------------------------------------------------------------------- 302
O4B-G-4
P4T-G-14
BAKING SYSTEM FOR EAST VACUUM VESSEL ------------------------------------------------------------------------------------------ 303
P4T-G-18
LASER MEGAJOULES CRYOGENIC TARGET DEVICES --------------------------------------------------------------------------------- 304
P4T-G-29
IRRADIATION TESTS ON WATER HYDRAULIC COMPONENTS --------------------------------------------------------------------- 305
P4T-G-53
EXPERIMENTAL RESULT OF THE LASER IN VESSEL VIEWING AND RANGING SYSTEM (IVVS) FOR ITER ---------- 306
P4T-G-56
ITER VACUUM VESSEL SECTOR MANUFACTURING DEVELOPMENT IN EUROPE ------------------------------------------- 307
P4T-G-71
STRUCTURAL UPGRADE OF IN-VESSEL CONTROL COIL ON DIII D*-------------------------------------------------------------- 308
P4T-G-89
MANUFACTURING OF CRYOSTAT FOR EAST SUPERCONDUCTING TOKAMAK ---------------------------------------------- 309
P4T-G-117
SPECIAL BLANKET DESIGN IN THE NB REGION OF ITER ----------------------------------------------------------------------------- 310
P4T-G-137
USE OF ELECTRONIC AND OPTOELECTRONIC INDUSTRIAL SYSTEMS FOR MAINTENANCE TOOLS OF ITER
FUSION EXPERIMENTAL REACTOR ---------------------------------------------------------------------------------------------------------- 311
P4T-G-199
NON-DESTRUCTIVE TESTING OF BONDED STRUCTURES FOR PLASMA FACING COMPONENTS ----------------------- 312
P4T-G-240
VERTICAL DIPLACEMENT EVENTS SIMULATIONS FOR TOKAMAK PLASMAS ----------------------------------------------- 313
P4T-G-249
DESIGN OF THE ITER HOT CELL BUILDING WITHDRAWN --------------------------------------------------------------------------- 314
P4T-G-264
RECENT DEVELOPMENTS TOWARDS ITER 2001 DIVERTOR MAINTENANCE -------------------------------------------------- 315
P4T-G-293
MANAGEMENT OF A WATER LEAK ON ACTIVELY COOLED FUSION DEVICES ---------------------------------------------- 316
P4T-G-296
DESIGN PROGRESS OF THE ITER VACUUM VESSEL AND PORTS ------------------------------------------------------------------ 317
P4T-G-348
1200 MM BORE VOLTAGE BREAK OF THE NB DUCT FOR KSTAR ------------------------------------------------------------------ 318
P4T-G-355
O4B-G-355
MANUFACTURE OF THE PLASMA VESSEL AND THE PORTS FOR WENDELSTEIN 7-X -------------------------------- 319
P4T-G-360
DYNAMIC IDENTIFICATION OF THE HYDRAULIC ITER MAESTRO MANIPULATOR - RELEVANCE FOR
MONITORING ---------------------------------------------------------------------------------------------------------------------------------------- 320
P4T-G-361
GENERIC CONTROL SYSTEM DESIGN FOR THE CASSETTE MULTIFUNCTION MOVER AND OTHER ITER REMOTE
HANDLING EQUIPMENT ------------------------------------------------------------------------------------------------------------------------- 321
P4T-G-374
ANALYSES OF THE ITER VACUUM VESSEL WITH THE USE OF A NEW MODELLING TECHNIQUE --------------------- 322
P4T-G-389
ITER ARTICULATED INSPECTION ARM (AIA): GEOMETRIC CALIBRATION ISSUES OF A LONG-REACH FLEXIBLE
ROBOT.------------------------------------------------------------------------------------------------------------------------------------------------- 323
P4T-G-393
ITER ARTICULATED INSPECTION ARM (AIA) : R&D PROGRESS ON VACUUM AND TEMPERATURE TECHNOLOGY
FOR REMOTE HANDLING. ----------------------------------------------------------------------------------------------------------------------- 324
P4T-G-404
ASSESSMENT OF A COOPERATIVE MAINTENANCE SCHEME FOR ITER DIVERTOR COOLING PIPE ------------------- 325
P4T-G-422
RF TESTS OF THE ELECTRICAL INSULATIONS FOR THE TOROIDAL STRUCTURES OF RFX ------------------------------ 326
P4T-G-435
OPERATIONAL EXPERIENCE FEEDBACK IN JET REMOTE HANDLING ----------------------------------------------------------- 327
P4T-G-509
IGNITOR PLASMA CHAMBER STRUCTURAL DESIGN WITH DYNAMIC LOADS DUE TO PLASMA DISRUPTION
EVENT -------------------------------------------------------------------------------------------------------------------------------------------------- 328
13
- H - Fuel Cycle. ----------------------------------------------------------------------------------------------------------------------------------------------------------- 329
P1C-H-17
ADVANCED PROCEDURES FOR TWO-STAGE REPETITIVE PELLET INJECTOR. ------------------------------------------------ 329
P1C-H-38
STUDIES OF PELLET DELIVERY AND SURVIVABILITY THROUGH CURVED GUIDE TUBES FOR FUSION FUELING
AND IMPLICATIONS FOR ITER ----------------------------------------------------------------------------------------------------------------- 330
P1C-H-93
O1B-H-93
PELLET INJECTORS FOR STEADY STATE FUELLING ------------------------------------------------------------------------------ 331
P1C-H-107
JET CONTRIBUTIONS TO THE ITER FUEL CYCLE ISSUES. ---------------------------------------------------------------------------- 332
P1C-H-153
COMPARISON OF MODELLING OF TRITIUM RELEASE FROM CERAMIC BREEDER MATERIALS ------------------------ 333
P1C-H-217
ASSESSMENT OF THE ITER DWELL EVACUATION ------------------------------------------------------------------------------------- 334
P1C-H-247
STRATEGY FOR DETERMINATION OF ITER IN-VESSEL TRITIUM INVENTORY ----------------------------------------------- 335
P1C-H-275
REQUIREMENTS AND SELECTION CRITERIA FOR THE MECHANICAL PUMPS FOR THE ITER TRITIUM PLANT --- 336
P1C-H-303
HIGH-POWER PULSED FLASHLAMP CLEANING OF CO-DEPOSITED HYDROCARBON FILMS FROM PLASMA
FACING COMPONENTS --------------------------------------------------------------------------------------------------------------------------- 337
P1C-H-358
GAS PUFFING BY MOLECULAR BEAM INJECTION IN ADITYA TOKAMAK ----------------------------------------------------- 338
P1C-H-438
INFLUENCE OF DEUTERIUM ON THE DESIGN OF THE JET WATER DETRITIATION SYSTEM ----------------------------- 339
P1C-H-441
EXPERIMENTAL VALIDATION OF A METHOD FOR PERFORMANCE MONITORING OF THE FRONT-END
PERMEATORS IN THE TEP SYSTEM OF ITER ---------------------------------------------------------------------------------------------- 340
P1C-H-461
PROTECTION OF THE PRIMARY CIRCUITS AND EFFECT ON THE DESIGN OF THE INNER DEUTERIUM / TRITIUM
FUEL CYCLE OF ITER ----------------------------------------------------------------------------------------------------------------------------- 341
P1C-H-467
EVALUATION OF SUPER CRITICAL HELIUM AS A COOLANT FOR DIII-D TYPE CRYOCONDENSATION -------------- 342
14
- I - Materials Technology and Breeding Blankets. --------------------------------------------------------------------------------------------------------------- 343
P1C-I-1
USE OF THE SPIRAL 2 FACILITY FOR MATERIAL IRRADIATIONS WITH 14 MEV ENERGY NEUTRONS --------------- 343
P1C-I-10
SCIENTIFIC AND TECHNICAL FOUNDATIONS AND TECHNOLOGIES OF REDUCTION OF MHD-RESISTANCE OF
DUCTS WITH HEAVY LIQUID METAL COOLANTS IN MAGNETIC FIELD OF BLANKET AND DIVERTER OF
TOKAMAK -------------------------------------------------------------------------------------------------------------------------------------------- 344
P1C-I-26
EFFECT OF UNDERSIZED SOLUTE ATOMS ON MICROSTRUCTURE CHANGE -------------------------------------------------- 345
P1C-I-39
RADIATION INDUCED CONDUCTIVITY AND SURFACE ELECTRICAL DEGRADATION OF PLASMA SPRAYED
SPINEL FOR NBI SYSTEMS ---------------------------------------------------------------------------------------------------------------------- 346
P1C-I-43
BLANKET MANUFACTORING TECHNOLOGIES : THERMOMECHANICAL TESTS ON HCLL BLANKET MOCKS UP 347
P1C-I-58
HIGH ENERGY PROTON DEGRADATION IN KU1 QUARTZ GLASS ----------------------------------------------------------------- 348
P1C-I-85
EXPERIMENTAL STUDY OF LITHIUM MHD FLOW IN SLOTTED CHANNEL FROM V-4TI-4CR ALLOY ----------------- 349
P1C-I-88
A NEUTRONIC INVESTIGATION OF HE-COOLED LI-BREEDER BLANKETS FOR FUSION POWER REACTOR --------- 350
P1C-I-96
MICROSTRUCTURAL CHARACTERISATION OF EUROFER-ODS RAFM STEEL IN THE NORMALIZED AND
TEMPERED CONDITION AND AFTER THERMAL AGING IN SIMULATED FUSION CONDITIONS ------------------------- 351
P1C-I-102
NON-DESTRUCTIVE ANALYSIS OF MINIATURIZED FUSION MATERIALS SAMPLES AND IRRADIATION CAPSULES
BY X RAY MICRO-TOMOGRAPHY ------------------------------------------------------------------------------------------------------------ 352
P1C-I-108
INNER STRUCTURES OF COMPRESSED PEBBLE BEDS DETERMINED BY X-RAY TOMOGRAPHY----------------------- 353
P1C-I-109
THERMAL CREEP OF BERYLLIUM PEBBLE BEDS --------------------------------------------------------------------------------------- 354
P1C-I-110
THERMAL CREEP BEHAVIOR OF THE EUROFER97 RAFM STEEL AND TWO EUROPEAN ODS-EUROFER97 STEELS
------------------------------------------------------------------------------------------------------------------------------------------------------------ 355
P1C-I-122
SEGREGATED VOID SWELLING IN A HETEROGENEOUS MATERIAL: IMPLICATIONS FOR FUSION MATERIALS-- 356
P1C-I-128
THERMOCHEMISTRY OF LI-TITANATES CERAMICS IN REDUCING ENVIRONMENTS -------------------------------------- 357
P1C-I-129
MOLECULAR DYNAMICS SIMULATIONS OF DEFECT PRODUCTION DURING IRRADIATION IN SILICA GLASS --- 358
P1C-I-130
KINETICS OF LI DEPLETED LI2TIO3 REACTION WITH H2 ADDED TO AR PURGE GAS -------------------------------------- 359
P1C-I-134
VITAMIN-J/COVA/EFF-3 CROSS-SECTION COVARIANCE MATRIX LIBRARY AND ITS USE TO ANALYSE
BENCHMARK EXPERIMENTS IN SINBAD DATABASE ---------------------------------------------------------------------------------- 360
P1C-I-140
IN-SITU FORMATION AND CHEMICAL STABILITY OF ER2O3 COATING ON V-4CR-4TI IN LIQUID LITHIUM -------- 361
P1C-I-141
PHYSICO-CHEMICAL PROPERTIES OF AND HYDROGEN ISOTOPE BEHAVIORS IN LITHIUM-TIN ALLOY AS A
LIQUID BREEDER FOR FUSION REACTOR ------------------------------------------------------------------------------------------------- 362
P1C-I-143
INTEGRAL EXPERIMENT ON BERYLLIUM WITH D-T NEUTRONS FOR VERIFICATION OF TRITIUM BREEDING --- 363
P1C-I-147
CREEP STRENGTH OF REDUCED ACTIVATION FERRITIC/MARTENSITIC STEEL EUROFER'97 --------------------------- 364
P1C-I-150
REACTION OF TITANIUM BERYLLIDE ------------------------------------------------------------------------------------------------------ 365
P1C-I-158
INTEGRAL BENCHMARK EXPERIMENTS ON VANADIUM SPHERES WITH A CENTRAL 14-MEV NEUTRON SOURCE
AND INSIDE A SPHERICAL CRITICAL ASSEMBLY -------------------------------------------------------------------------------------- 366
P1C-I-163
O1A-I-163
PRESENT DEVELOPMENT STATUS OF EUROFER AND ODS FOR APPLICATION IN BLANKET CONCEPTS ---- 367
P1C-I-164
MICROSTRUCTURAL INVESTIGATION, USING SMALL ANGLE NEUTRON SCATTERING, OF NEUTRON
IRRADIATED EUROFER 97 STEEL ------------------------------------------------------------------------------------------------------------- 368
P1C-I-168
EFFECT ON IMPACT TOUGHNESS OF REDUCED OXYGEN CONTENT IN 316 STEEL POWDER JOINED TO 316 STEEL
BY LOW TEMPERATURE HIP ------------------------------------------------------------------------------------------------------------------- 369
P1C-I-178
ENVIRONMENTAL ASSISTED CRACKING OF EUROFER 97 IN WATER AND PB-LI -------------------------------------------- 370
P1C-I-179
MEASUREMENT AND ANALYSIS OF RADIOACTIVITY INDUCED IN YTTRIUM AND LEAD IN FUSION PEAK
NEUTRON FIELD ------------------------------------------------------------------------------------------------------------------------------------ 371
P1C-I-183
EVALUATION OF NUCLEAR HEATING, TRITIUM BREEDING AND SHIELDING EFFICIENCY OF THE DEMO HCLL
BREEDER BLANKET ------------------------------------------------------------------------------------------------------------------------------- 372
P1C-I-189
HYDROGEN EFFECTS ON THE TENSILE AND FATIGUE PROPERTIES OF EUROFER 97 -------------------------------------- 373
P1C-I-191
THE HELIUM COOLED LITHIUM LEAD BLANKET TEST PROPOSAL IN ITER AND REQUIREMENTS ON TEST
BLANKET MODULES INSTRUMENTATION ------------------------------------------------------------------------------------------------ 374
P1C-I-196
NUMERICAL AND EXPERIMENTAL STUDY ON TIME-DEPENDENT THERMOMCHANIC DEFORMATION OF
CERAMIC BREEDER PEBBLE BEDS ----------------------------------------------------------------------------------------------------------- 375
P1C-I-197
HYDROGEN ISOTOPE DISTRIBUTIONS AND RETENTION IN THE INNER DIVERTOR TILE OF JT-60U ------------------ 376
P1C-I-200
CRYSTAL STRUCTURE OF LI2TIO3 WITH SOME DIFFERENT OXIDE ADDITIVES --------------------------------------------- 377
15
P1C-I-205
EVALUATION OF INSULATING PROPERTY OF CERAMIC MATERIALS FOR V/LI BLANKET SYSTEM UNDER
FISSION REACTOR IRRADIATION ------------------------------------------------------------------------------------------------------------- 378
P1C-I-214
EVALUATION OF HYDROGEN ISOTOPE RETENTION IN BE12TI AS NEUTRON MULTIPLIER OF FUSION REACTOR
------------------------------------------------------------------------------------------------------------------------------------------------------------ 379
P1C-I-215
MECHANICAL PROPERTIES OF WELDAMENT USING IRRADIATED STAINLESS STEEL FOR BLANKET -------------- 380
P1C-I-224
AB-INITIO VALUES OF THE HE SIEVERT´S CONSTANT IN LIQUID LI ------------------------------------------------------------- 381
P1C-I-237
OUT-OF-PILE TRITIUM RELEASE PROPERTY CORRELATIONS FOR LI-DEPLETED LI2TIO3 AND LI4TI5O12
CERAMICS. EFFECTS OF REDUCTION-ANNEALING TREATMENTS --------------------------------------------------------------- 382
P1T-I-238
MAGNETOHYDRODYNAMIC PRESSURE-DRIVEN FLOWS IN THE HCLL BLANKET ------------------------------------------ 383
P1T-I-242
LIQUID LITHIUM AS THE COOLANT OF THE IFMIF LOOP ---------------------------------------------------------------------------- 384
P1T-I-252
HCLL TBM FOR ITER – DESIGN STUDIES --------------------------------------------------------------------------------------------------- 385
P1T-I-254
INTERNATIONAL COMPARISON OF MEASURING TECHNIQUES OF TRITIUM PRODUCTION FOR FUSION
NEUTRONICS EXPERIMENTS ------------------------------------------------------------------------------------------------------------------- 386
P1T-I-257
PEBBLE BED THERMAL-MECHANICAL THEORETICAL MODEL: APPLICATION AT THE GEOMETRY OF TEST
BLANKET MODULE OF ITER-FEAT NUCLEAR FUSION REACTOR ----------------------------------------------------------------- 387
P1T-I-269
BEHAVIOUR OF TRITIUM IN BREEDING BLANKET MATERIALS ------------------------------------------------------------------- 388
P1T-I-276
MUTUAL CORROSION OF EUROFER97 AND THE BLANKET CERAMIC MATERIALS ----------------------------------------- 389
P1T-I-287
AUTOMATIC GENERATION OF A JET 3D NEUTRONICS MODEL FROM CAD GEOMETRY DATA FOR MONTE CARLO
CALCULATIONS ------------------------------------------------------------------------------------------------------------------------------------ 390
P1T-I-289
PERFORMANCE OF A HYDROGEN SENSOR IN PB-16LI WITHDRAWN ----------------------------------------------------------- 391
P1T-I-292
THE CHARACTERIZATION AND STRESS ANALYSIS ON VACUUM PLASMA SPRAYING TUNGSTEN COATINGS --- 392
P1T-I-295
SOME FEATURES OF BERYLLIUM CORROSION BEHAVIOUR IN BE-LIQUID LI-V4 TI 4 CR ALLOY SYSTEM -------- 393
P1C-I-302
BERYLLIUM AS BLANKET MATERIAL: PECULIARITIES OF RADIATION DAMAGE UNDER HIGH DOSE
NEUTRON IRRADIATION WITHDTRAWN ------------------------------------------------------------------------------------------------- 394
O1A-I-302
P1T-I-307
TRITIUM BREEDING EXPERIMENTS WITH BLANKET MOCK-UPS CONTAINING 6LI-ENRICHED LITHIUM TITANATE
AND BERYLLIUM IRRADIATED WITH DT NEUTRONS --------------------------------------------------------------------------------- 395
P1T-I-308
EFFECTS OF GELATION AND SINTERING CONDITIONS ON GRANULATION OF LI2TIO3 PEBBLES FROM LI-TI
COMPLEX SOLUTION ----------------------------------------------------------------------------------------------------------------------------- 396
P1T-I-309
FUSION-DRIVEN HYBRID SYSTEM WITH ITER MODEL ------------------------------------------------------------------------------- 397
P1T-I-313
SURFACE WAVE ON HIGH SPEED LIQUID LITHIUM FLOW FOR IFMIF ----------------------------------------------------------- 398
P1T-I-316
MEASUREMENT OF ENERGETIC CHARGED PARTICLES PRODUCED IN FUSION MATERIALS WITH 14 MEV
NEUTRON IRRADIATION------------------------------------------------------------------------------------------------------------------------- 399
P1T-I-318
STRUCTURAL ANALYSIS FOR THE GAS-COOLED HIGH FLUX TEST MODULE OF IFMIF ---------------------------------- 400
P1T-I-319
THERMAL AND THERMAL-STRESS ANALYSES OF IFMIF LIQUID LITHIUM TARGET ASSEMBLY ---------------------- 401
P1T-I-321
THERMAL HYDRAULIC ANALYSIS OF FDS-II LIPB BREEDER BLANKET -------------------------------------------------------- 402
P1T-I-323
THERMAL DESORPTION BEHAVIOR OF HYDROGEN ISOTOPES INTERACTING WITH RADIATION DEFECTS IN LI2O
------------------------------------------------------------------------------------------------------------------------------------------------------------ 403
P1T-I-325
PRESENT STATUS OF BERYLLIDE STUDY FOR FUSION AND APPLICATION DEVELOPMENT IN JAPAN -------------- 404
P1T-I-327
EFFECTS OF IRRADIATION ON MECHANICAL PROPERTIES OF HIP-BONDED F82H STEEL-------------------------------- 405
P1T-I-329
ACTIVATION OF EUROFER IN AN IFMIF-LIKE NEUTRON FIELD ------------------------------------------------------------------- 406
P1T-I-331
JOINING OF CFC TO COPPER FOR ITER DIVERTOR ------------------------------------------------------------------------------------- 407
P1T-I-334
DEVELOPMENT OF RF-INPUT COUPLER WITH A MULTI-LOOP ANTENNA FOR RFQ LINAC IN IFMIF PROJECT---- 408
P1T-I-335
IN-SITU IN-REACTOR TESTING OF FUSION MATERIALS AND COMPONENTS ------------------------------------------------- 409
P1T-I-339
NEUTRONICS AND ACTIVATION CHARACTERISTICS OF THE INTERNATIONAL FUSION MATERIAL IRRADIATION
FACILITY ---------------------------------------------------------------------------------------------------------------------------------------------- 410
P1T-I-340
DESIGN, MANUFACTURING AND TESTING OF THE IFMIF LITHIUM TARGET BAYONET CONCEPT BACKPLATE. 411
P1T-I-365
EFFECTIVE THERMAL CONDUCTIVITY OF A COMPRESSED LI2TIO3 PEBBLE BED ------------------------------------------ 412
P1T-I-368
CONCEPTUAL DESIGN OF THE BLANKET MECHANICAL ATTACHMENT FOR THE HELIUM-COOLED LITHIUMLEAD REACTOR ------------------------------------------------------------------------------------------------------------------------------------- 413
P1T-I-378
DEVELOPMENT OF EXPERIMANTAL DEVICES FOR IN-REACTOR MECHANICAL TESTS ---------------------------------- 414
P1T-I-388
NEW MODULAR CONCEPT FOR THE HELIUM COOLED PEBBLE BED TEST BLANKET MODULE FOR ITER ---------- 415
P1T-I-391
THE TEMPERATURE DEPENDENCE OF STRAIN-RATE EFFECT ON TENSILE STRENGTH OF MO-ALLOYS ----------- 416
P1T-I-397
MECHANICAL AND THERMAL PROPERTIES OF SIC/SIC COMPOSITES IRRADIATED WITH NEUTRONS AT HIGH
TEMPERATURES ------------------------------------------------------------------------------------------------------------------------------------ 417
16
P1T-I-402
THERMAL-HYDRAULIC ANALYSIS AND OPTIMISATION OF THE BREEDER UNIT FOR THE EU HELIUM COOLED
PEBBLE BED BLANKET --------------------------------------------------------------------------------------------------------------------------- 418
P1T-I-407
INFLUENCE OF NEUTRON IRRADIATION ON TOUGHNESS AND R-CURVE BEHAVIOUR OF SIC/SIC ------------------- 419
P1T-I-409
IN-VESSEL INTEGRATION OF THE MODULAR EU HELIUM COOLED PEBBLE BED BLANKET IN A DEMORELEVANT TOKAMAK GEOMETRY ---------------------------------------------------------------------------------------------------------- 420
P1T-I-411
DEVELOPMENT AND FABRICATION ASPECTS REGARDING TUNGSTEN COMPONENTS FOR A HE-COOLED
DIVERTOR -------------------------------------------------------------------------------------------------------------------------------------------- 421
P1T-I-415
SLIP INFLILTRATION AND DENSIFICATION OF POROUS SICF/SIC PREFORMS USING SIC NANOPOWDERS -------- 422
P1T-I-416
DESIGN OF FDS DEMO BLANKETS AND TEST BLANKET MODULE PROPOSED FOR ITER --------------------------------- 423
P1T-I-418
STATUS OF THE HFR PETTEN HIGH DOSE IRRADIATION ----------------------------------------------------------------------------- 424
P1T-I-419
HYDROGEN ISOTOPES BEHAVIOR ON LI2TIO3 UNDER VARIED SURFACE CONDITION ----------------------------------- 425
P1T-I-425
DEFORMATION BEHAVIOUR OF COPPER UNDER IN-REACTOR UNIAXIAL TENSILE TESTS------------------------------ 426
P1T-I-430
A HIGH FLUENCE IRRADIATION OF CERAMIC BREEDER MATERIALS IN HFR PETTEN, MATERIALS
CHARACTERISATION AND TEST MATRIX. ------------------------------------------------------------------------------------------------ 427
P1T-I-432
CHARACTERIZATION AND STABILITY STUDIES OF TITANIUM BERYLLIDES ------------------------------------------------ 428
P1T-I-434
IN-SITU BONDING OF SIC/SIC BY CONTROLLED SHS COMBUSTION ------------------------------------------------------------- 429
P1T-I-440
NEUTRONIC DESIGN OPTIMISATION OF MODULAR HCPB BLANKETS FOR FUSION POWER REACTORS ------------ 430
P1C-I-444
O1A-I-444
THE EUROPEAN BREEDING BLANKETS DEVELOPMENT AND THE TEST STRATEGY IN ITER -------------------- 431
P1T-I-450
FABRICATION OF YTTRIUM OXIDE AND ERBIUM OXIDE COATINGS BY PVD METHODS--------------------------------- 432
P1T-I-454
ON THE HYPERPOROUS NON-LINEAR ELASTICITY MODEL FOR FUSION-RELEVANT PEBBLE BEDS ----------------- 433
P1T-I-464
INFLUENCE OF HEATING TREATMENT AND MICROSTRUCTURE ON TRITIUM DESORPTION KINETIC -------------- 434
P1T-I-465
TRANSMUTATION AND ACTIVATION OF RUSSIAN STRUCTURAL MATERIALS FOR FUSION REACTORS IN
NEUTRON SPECTRA OF FISSION AND FUSION REACTORS --------------------------------------------------------------------------- 435
P1T-I-470
ON THE NUCLEAR RESPONSE OF THE HELIUM-COOLED LITHIUM LEAD TEST BLANKET MODULE IN ITER ------ 436
P1C-I-472
IN-PILE PERFORMANCE OF THE CERAMIC BREEDER PEBBLE-BED ASSEMBLIES FOR THE HCPB BLANKET
CONCEPT --------------------------------------------------------------------------------------------------------------------------------------------- 437
O1A-I-472
P1T-I-493
PRODUCTION OF LOW ACTIVATION V-(4-5)TI-(4-5)CR ALLOYS FOR FUSION REACTOR APPLICATIONS.
WITHDRAWN ---------------------------------------------------------------------------------------------------------------------------------------- 438
P1T-I-494
HEAT RESISTANT RAFMS RUSFER-EK-181 FOR FUSION AND FAST BREEDER REACTORS APPLICATIONS
WITHDRAWN ---------------------------------------------------------------------------------------------------------------------------------------- 439
P1T-I-496
GETTERING OF NITROGEN IN LIQUID LITHIUM ----------------------------------------------------------------------------------------- 440
P1C-I-497
O3B-I-497
PROSPECTIVE TESTING PROGRAMME FOR IFMIF --------------------------------------------------------------------------------- 441
P1T-I-498
NEUTRONIC OPTIMIZATION ANALYSIS OF FDS-‡U LIPB BREEDER BLANKET ----------------------------------------------- 442
P1T-I-516
PRODUCTION AND THERMAL STABILITY OF BERYLLIUM WITH FINE GRAIN STRUCTURE TO IMPROVE TRITIUM
RELEASE DURING NEUTRON IRRADIATION ---------------------------------------------------------------------------------------------- 443
P1T-I-520
ITER MATERIALS PROPERTIES DATA ------------------------------------------------------------------------------------------------------- 444
P1T-I-535
MIXED MHD CONVECTION AND TRITIUM TRANSPORT IN FUSION-RELEVANT CONFIGURATIONS ------------------ 445
P1T-I-536
OPTIMIZATION OF REDUCED ACTIVATION MARTENSITIC STEEL F82H FOR DEMO BREEDING BLANKET --------- 446
P1T-I-537
EFFECT OF TEMPERATURE CHANGE ON THE IRRADIATION HARDENING OF MARTENSITIC AND AUSTENITIC
STEELS IRRADIATED TO 1.5 DPA IN JMTR ------------------------------------------------------------------------------------------------- 447
P1T-I-538
OXIDE DISPERSION STRENGTHENING STEELS R&D FOR WATER-COOLING FUSION BLANKET SYSTEM ----------- 448
17
- J - Power Plants, Safety and Environment, Socio-economics.---------------------------------------------------------------------------------------------- 449
P4T-J-21
LOW LEVEL CLEANING OF A FUSION TARGET CHAMBER -------------------------------------------------------------------------- 449
P4T-J-44
THE EVITA PROGRAMME: EXPERIMENTAL AND NUMERICAL SIMULATION OF A FLUID INGRESS IN THE
CRYOSTAT OF A WATER-COOLED FUSION REACTOR --------------------------------------------------------------------------------- 450
P4T-J-45
CORROSION OF FUSION-SPECIFIC WASTE MATERIALS ------------------------------------------------------------------------------ 451
P4T-J-63
MATERIALS ACTIVATION INDUCED BY HIGH ENERGY NEUTRONS: A COMPARISON OF ANITA-IEAF
CALCULATION WITH MEASUREMENTS FROM THE KARLSRUHE ISOCHRONOUS CYCLOTRON ----------------------- 452
P4T-J-65
EXPERIMENTAL FUSION MATERIAL PHOTON AND ELECTRON DECAY HEAT MEASUREMENTS: ITS USE FOR
ACTIVATION CODES VALIDATION ----------------------------------------------------------------------------------------------------------- 453
P4T-J-70
SAFETY ANALYSIS FOR ITER LICENSING-------------------------------------------------------------------------------------------------- 454
P4T-J-83
VALIDATION OF THE ECART CODE FOR THE SAFETY ANALYSIS OF FUSION REACTORS -------------------------------- 455
P4T-J-86
RADIOACTIVE WASTE MANAGEMENT FOR THE IGNITOR FUSION EXPERIMENT ------------------------------------------- 456
P4T-J-124
3D-ANALYSIS OF AN ITER ACCIDENT SCENARIO --------------------------------------------------------------------------------------- 457
P4T-J-127
CATEGORISATION OF ACTIVATED MATERIAL FROM FUSION POWER REACTORS AND ACCEPTABILITY FOR
FINAL DISPOSAL ----------------------------------------------------------------------------------------------------------------------------------- 458
P4T-J-138
DUST IN ITER: R&D NEEDS FOR SAFETY COMPLIANCE ------------------------------------------------------------------------------ 459
P4T-J-139
RADIOACTIVE WASTE FROM A D-HE3 REACTOR --------------------------------------------------------------------------------------- 460
P4T-J-148
FIBER OPTIC SENSORS NETWORKS FOR ENVIRONMENTAL AND SAFETY MONITORING OF FUSION REACTORS 461
P4T-J-151
O4B-J-151
THE ECONOMIC VIABILITY OF FUSION POWER ------------------------------------------------------------------------------------- 462
P4T-J-170
ITER DIVERTOR EX-VESSEL PIPE BREAK -------------------------------------------------------------------------------------------------- 463
P4T-J-177
ENVIRONMENTAL RELEASE TARGETS FOR FUSION POWER PLANTS ----------------------------------------------------------- 464
P4T-J-193
DYNAMIC ASSESSMENTS OF CHAMBER AND WALL RESPONSE TO TARGET IMPLOSION IN INERTIAL FUSION
REACTORS -------------------------------------------------------------------------------------------------------------------------------------------- 465
P4T-J-216
CONSEQUENCE CALCULATIONS FOR PPCS BOUNDING ACCIDENTS ------------------------------------------------------------ 466
P4T-J-222
COMPONENT FAILURE DATA COLLECTION AND ANALYSIS FROM JET AND TLK OPERATING EXPERIENCE ----- 467
P4T-J-232
COLLECTION AND ANALYSIS OF OCCUPATIONAL RADIATION EXPOSURE DATA RELATED TO JET OPERATIONS
------------------------------------------------------------------------------------------------------------------------------------------------------------ 468
P4T-J-244
TFTR OCCUPATIONAL RADIATION EXPOSURE DATA COLLECTION AND ANALYSIS -------------------------------------- 469
P4T-J-260
FACTORS AFFECTING THE INHALATION DOSE FROM TRITIATED DUST AND FLAKES------------------------------------ 470
P4T-J-271
O4B-J-271
THE EUROPEAN POWER PLANT CONCEPTUAL STUDY --------------------------------------------------------------------------- 471
P4T-J-273
INTRA ANALYSIS OF WET BYPASS TRANSIENTS INCLUDING TRITIUM -------------------------------------------------------- 472
P4T-J-317
ACCESSIBILITY EVALUATION OF THE IFMIF LIQUID LITHIUM LOOP CONSIDERING ACTIVATED
EROSION/CORROSION MATERIALS DEPOSITION --------------------------------------------------------------------------------------- 473
P4T-J-336
AVAILABILITY OF LITHIUM IN THE CONTEXT OF FUTURE D-T FUSION REACTORS --------------------------------------- 474
P4T-J-342
EFFECT OF ACTIVATION CROSS-SECTION UNCERTAINTIES IN SELECTING STEELS FOR THE HYLIFE-II CHAMBER
TO SUCCESSFUL WASTE MANAGEMENT -------------------------------------------------------------------------------------------------- 475
P4T-J-357
EVALUATION OF FUSION STUDY FROM SOCIO-ECONOMIC ASPECTS ----------------------------------------------------------- 476
P4T-J-373
PROGRESS IN THE DEVELOPMENT OF A PIE-PIT FOR THE ITER TOKAMAK --------------------------------------------------- 477
P4T-J-380
GLOBAL ENERGY MODEL WITH FUSION -------------------------------------------------------------------------------------------------- 478
P4T-J-400
DUST EXPLOSION HAZARD IN ITER: EXPLOSION INDICES OF FINE GRAPHITE AND TUNGSTEN DUSTS AND THEIR
MIXTURES -------------------------------------------------------------------------------------------------------------------------------------------- 479
P4T-J-401
ECONOMIC ANALYSIS OF FDS FUSION POWER REACTORS ------------------------------------------------------------------------- 480
P4T-J-414
RELIABILITY ANALYSIS OF BLANKET MODULES OF FDS --------------------------------------------------------------------------- 481
P4T-J-421
FUSION SAFETY STUDIES IN RUSSIA IN 2003. -------------------------------------------------------------------------------------------- 482
P4T-J-423
CORE CONCEPTUAL DESIGN OF FDS FUSION POWER REACTORS ---------------------------------------------------------------- 483
P4T-J-452
NEUTRON ACTIVATION AND DOSE RATES MINIMIZATION ON LASER MÉGAJOULE (LMJ) FACILITY --------------- 484
P4T-J-524
DESIGN EARTHQUAKES FOR ITER AT CADARACHE ----------------------------------------------------------------------------------- 485
P4T-J-525
METHODOLOGY FOR REFERENCE ACCIDENTS DEFINITION FOR ITER --------------------------------------------------------- 486
P4T-J-527
FIRE RISK ANALYSIS IN ITER TRITIUM BUILDING ------------------------------------------------------------------------------------- 487
18
P4T-J-529
CHEMICAL RISK STUDIES INCLUDING BERYLLIUM AND CHEMICAL ZONING ----------------------------------------------- 488
P4T-J-530
PROGRESS IN LICENSING ITER IN CADARACHE----------------------------------------------------------------------------------------- 489
P4T-J-532
ALARA APPLIED TO ITER DESIGN. RADIOPROTECTION AND ZONING APPROACH ----------------------------------------- 490
19
- K - Transfer of Technology. ------------------------------------------------------------------------------------------------------------------------------------------ 491
P1C-K-131
100 KV SOLID-STATE SWITCH FOR FUSION HEATING SYSTEMS ------------------------------------------------------------------ 491
P1C-K-443
O1B-K-443
OVERVIEW OF CRYOGENIC REFRIGERATION SYSTEMS FOR THE THERMONUCLEAR FUSION ---------------- 492
P1C-K-499
DEVELOPMENT & APPLICATION OF MCNP AUTO-MODELING TOOL : MCAM 3.0 -------------------------------------------- 493
20
- A - CURRENT AND NEXT STEP DEVICES
P3C-A-11
SELECTION OF DESIGN SOLUTIONS AND FABRICATION METHODS
AND SUPPORTING R&D FOR PROCUREMENT OF ITER VESSEL AND
FW/BLANKET
IOKI, KIMIHIRO, THE ITER INTERNATIONAL TEAM AND PARTICIPANT TEAMS
ITER Garching JWS, Boltzmannstrabe 2, 85748 Garching, Germany
The ITER project has started preparation of Procurement Specification Documents for key components. The
design of the ITER vacuum vessel (VV) and first wall (FW)/blanket has progressed by selecting design
solutions, and R&D results are providing the basis for selection of design solutions and fabrication methods.
The VV design has progressed in many aspects, such as an independent cooling configuration in the VV field
joint regions, 9 lower ports instead of 18, a single wall structure for the upper and equatorial ports except the
NB ports, and the vacuum vessel gravity support located below the lower ports. Double curvature pressing is
now selected instead of facet shape welding for inner and outer shells in the upper and lower inboard regions
to improve the fabrication and NDT process. By this selection, very short distances between neighbouring
welds can be avoided. A challenging UT R&D program is also going on to achieve acceptable S/N ratio for
small-angle launching waves (20-30 deg.). Another approach is a combination of progressive PT and
conventional UT. Selection of the NDT method in critical areas will be made based on R&D results.
Regarding the FW/blanket system, the plasma facing surface of the FW has been defined to avoid protruding
the leading edges, especially in the inboard area. Separate FW panels are supported with a central beam, and
selection of a race-track shape cross-section for the central beam provides a more robust structure against halo
current EM loads and also leads to a new cooling configuration in the shield block, where the pressure drop is
significantly reduced to ~0.05 MPa. Detailed EM analysis has been performed by using a newly defined
plasma current quench scenario (40 ms linear decay and 25 ms exponential decay), and EM loads due to eddy
currents are reduced in the current design with deeper slits and lower steel/water ratio in the shield block. The
welding/cutting method of the FW central beam in the hot cell will be selected between YAG laser and TIG
welding/mechanical cutting, based on R&D results. For future higher performance operation, the possibility of
long pulse operation (3000/1000 s burn time in non-inductive/hybrid operation) and high fusion power
operation (700MW) have been assessed. Helium purge gas lines for the ITER breeding blanket have been
designed and analysed as a parameter of the tritium partial pressure in the range 1-50 Pa, and further testing is
proposed to select the parameter.
Corresponding Author:
IOKI, KIMIHIRO
iokik@itereu.de
Ioki, Kimihiro
21
- A - Current and Next Step Devices
P3C-A-16
THE PROTO-SPHERA LOAD ASSEMBLY
PAPASTERGIOU STAMOS, ALLADIO FRANCO MICOZZI PAOLO MANCUSO ALESSANDRO
c/0 ENEA, VIA E FERMI 45, FRASCATI 00044,ROMA
THE PROTO-SPHERA LOAD ASSEMBLY S Papastergiou, F Alladio, A Mancuso, P Micozzi Absract
PROTO-SPHERA is a proposed spherical torus where a hydrogen plasma arc, in a form of a screw pinch field
fed by electrodes , replaces the central conductor. This simply connected magnetic configuration, if fusion
relevant, might strongly simplify the design of a fusion reactor. The machine design philosophy, basic
geometry and operating conditions together with the major components like the vacuum vessel, water cooled
coils, electrodes, protection components, divertor etc will be analysed. The thermal and electromagnetic
behavior, the duty cycle as well as the predicted and permitted key stresses will be discussed in order to prove
that the design, construction and reliable operation of the machine are feasible as demonstrated in an
international workshop at ENEA-Frascati in March 2002. Finally reference should be made to the proposed
Multi-Pinch experiment, using the START vacuum vessel, to demonstrate the feasibility and stability of the
Proto-sfera configuration.
Corresponding Author:
PAPASTERGIOU STAMOS
papastergiou@frascati.enea.it
c/o ENEA ,VIA E FERMI 45 ,FRASCATI 00044, ROMA
22
- A - Current and Next Step Devices
P3C-A-72
OVERVIEW OF THE DIII–D PROGRAM AND CONSTRUCTION
PLANS*
PETERSEN, P.I., THE DIII-D TEAM
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
Selected also for oral presentation
O3B-A-72
The DIII-D tokamak is a mid size tokamak operating at reactor relevant parameters. Because of its size it is
relatively easy to modify the machine as required to test new ideas or theories. During the last few years
several new hardware items have been added to the DIII-D tokamak and improvements have been made to
others. The main additions in the last two years were the installation of the I-coil system and upgrades to the
electron cyclotron heating (ECH) system. In addition the fast wave system is being brought back into operation
after having been idle for three years. The I-coil system, which consists of 12 coils installed inside the DIII-D
vessel, is used to stabilize the resistive wall modes and to produce a stochastic edge, which has suppressed
edge localized modes (ELMs). ELMs can be detrimental to ITER, since they can erode the plasma facing
surfaces. The I-coils are powered by three switching power amplifying units, which together with a flexible
patch panel allow the I-coils to be operated in many different configurations. The ECH system has been
upgraded to six gyrotrons, which have been used to heat the plasma, modify the current profile and stabilize
the neoclassical tearing 3/2 and 2/1 modes. Three ECH launchers built by Princeton Plasma Physics
Laboratory are installed on the DIII-D tokamak and have the capability of changing the beam direction in both
toroidal and poloidal directions. Three additional gyrotrons have been ordered for the DIII-D program. They
are required for current profile control and stabilization of the NTMs. The gyrotrons are scheduled to be
installed during a 10–12 month facility enhancement period, which spans 2005–2006. At the same time a
modification is scheduled to be made to the lower divertor to make it pump double-null high triangularity
plasmas, which are important for studying advanced tokamak plasmas. One of the four neutral beam lines will
be rotated for counter injection, which will allow study of the quiescent double barrier mode with central corotation of the plasma and of the resistive wall mode with low rotation. *Work was supported by the U.S.
Department of Energy under DE-FC02-04ER54698.
Corresponding Author:
PETERSEN, P.I.
petersen@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
23
- A - Current and Next Step Devices
P3C-A-90
COMMISSIONING AND PRELIMINARY OPERATION OF THE HL-2A
TOKAMAK
LIU DEQUAN, LIU YONG , YAN JIANCHEN , CAO ZENG , YANG QINGWEI , ZHOU CAIPIN , LI
XIAODONG ,AND THE HL-2A TEAM
Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan,610041, P.R. China
HL-2A is a new operating tokamak in SWIP of China, it can be operated in double-null and single-null
divertor with closed configurations. The effect of the divertor on impurity behaviors, MHD instabilities,
transport, wall conditioning, divertor physics are key issues to study during the first step operation on HL-2A.
The construction of the HL-2A project had been finished in the fall of 2002, the first plasma was obtained in
the end of 2002. The improving of the vacuum system and other subsystems such as vessel inner pumping,
control system and power supply system had been carried out in 2003,the feedback control of the plasma
current and plasma position were used on both in limiter and divertor operations. Preliminary experiment with
limiter and single null divertor configurations were achieved in 2003. Primarily results of 168KA plasma
current , 920ms duration time and plasma linear average density of 1.7*10+13 cm-3 were obtained, impurity
especial the low Z impurity was clearly decreased during the divertor operation. During the operation,the
vacuum vessel was baked up to 115 C by hot water, glow discharge clearing was applied for approximate 120
hours with four electrodes , Ti metallic getters worked for about 10 hours in all.So far, the best limit vacuum
obtained is 4.6*10-6Pa on HL-2A.Cryogetter pump is very useful tool to absorb H2O in fusion device, but the
effect of a small capability pump used on HL-2A is not obvious,two great Cryogetter pumps will be used on
HL-2A in 2004, better vacuum will be gotten. Higher plasma parameters will be expected with a enhaned
power supply system in 2004.
Corresponding Author:
LIU DEQUAN
liudq@swip.ac.cn
Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan,610041, P.R. China
24
- A - Current and Next Step Devices
P3C-A-144
PLASMA PHYSICS BASIS AND OPERATIONS OF FUSION-DRIVEN
SUBCRITICAL SYSTEM
BIN WU,
The Fusion Driven Sub-critical System (FDS) is a sub-critical nuclear energy system drive by fusion neutron
source, which provides a feasible, safe, economic and highly efficient potential of disposing High Level Waste
(HLW) and produce fission nuclear fuel as a early application of fusion technology. The system includes a
tokamak as fusion neutron driver, a nuclear power system as blanket. Parameters of such kind reactor are
following. major radius 4m, minor radius 1m, plasma current 5.7MA, toroidal field 5.2T, Bootstrap current
fraction 0.90, Fusion power 143MW, Neutron wall loading 0.5MW/m2 . In this paper, an advanced plasma
configuration for FDS system has been proposed, which aims at high bata, high bootstrap current and good
confinement. The JSOLVER code has been used to getting equilibrium. Several different advance equilibrium
configurations have been proposed. Among these modes, the reverse shear mode is most attractive. In order to
determine the feasibility of tokamak operation, a transient simulation has been made which includes the
equilibrium, transport and plasma position shape control in FDS. A 1.5D equilibrium evolution code has been
used to make this simulation. The code is two-dimensional time dependent free boundary simulation code that
advances the MHD equations describing the transport time-scale evolution of axisymmetric tokamak plasma.
A detail plasma configuration evolution is obtained by this calculation. The simulation results confirm and
constrain the system projections.
Corresponding Author:
BIN WU
wubin@ipp.ac.cn
Institute of Plasma Physics, Chinese Academy of Sciences,P.O. Box 1126, Hefei, Anhui, 230031, China
25
- A - Current and Next Step Devices
P3C-A-184
THE WENDELSTEIN 7-X MECHANICAL STRUCTURE SUPPORT
ELEMENTS: TESTS AND DESIGN
GASPAROTTO, MAURIZIO, MR. SIMON-WEIDNER AND THE W7-X TEAM
Max-Planck-Institut für Plasmaphysik Wendelsteinstraße 1 17491 Greifswald Germany
The WENDELSTEIN 7-X Mechanical Structure Support Elements: Design and Tests M. Gasparotto, J.
Simon-Weidner and the W 7-X team Max-Planck-Institut für Plasmaphysik, Euratom Association, Teilinstitut
Greifswald, Wendelsteinstraße 1, D-17491Greifswald, Germany The stellarator WENDELSTEIN 7-X is in the
construction phase at IPP Geifswald, Germany. The main parameters are: average major radius 5.5 m, average
plasma radius 0.53 m, maximum magnetic field on the plasma axis 3.0 T, total weight 725t. The magnetic
system of the machine consists of 50 superconducting Non-Planar Coils (NPC), 20 superconducting Planar
Coils (PC), the Coil Support Structure and the Intercoil Support Structure (ISS). Each PC and NPC is
supported by the Coil Support Structure through two Coil Connection Elements (CCE) that must transmit
loads and moments up to 3.3 MN and 400 MNmm respectively. All components of the coil system are kept at
4K by liquid helium. The ISS consists of: (i) The Narrow Support Elements connecting adjacent NPC casings
in the inner region by sliding joints; (ii) The Lateral Supports connecting adjacent NPC casings at the outer
region by welded joints; (iii) The Planar Supports connecting the PC to the NPC by sliding joints. The CCE
and the ISS are critical components that should satisfy the following requirements: operate in high vacuum and
at cryogenic temperature, withstand high loads and moments and allow the assembly of the machine with high
accuracy while minimising the distortion due to the welded connections. A large R&D programme is in
progress to qualify the adopted solutions and to check the components which are most critical under operating
conditions. The CCE based on inconel bolted connections, is tested in scale 1:1 at 77 K applying the maximum
load and moment; the Narrow Support is tested at R.T. at the maximum load (150t) simulating tilting and
sliding movements while samples of the material are tested under vacuum and at cryogenic temperature
applying a sliding movement and a pressure of about 500 MPa. An R&D programme based on analytical
simulation and experimental tests is in progress to optimize the weld sequence during the assembly of the
magnetic system of W7-X. The paper will report the main design features of the CCE and ISS and results of
tests carried out to qualify materials and critical components.
Corresponding Author:
GASPAROTTO, MAURIZIO
maurizio.gasparotto@ipp.mpg.de
Wendelsteinstraße 1,D- 17491 Greifswald, Germany
26
- A - Current and Next Step Devices
P3C-A-203
LEVITATION EXPERIMENTS OF A HIGH TEMPERATURE
SUPERCONDUCTOR COIL IN THE INTERNAL COIL DEVICE MINIRT
MORIKAWA JUNJI, OGAWA YUICHI (1) OHKUNI KOTARO (1) YAMAKOSHI SHIGEO (2) GOTO TAKUYA (2)
MITO TOSHIYUKI (3) YANAGI NAGATO (3) IWAKUMA MASATAKA (4) UEDE TOSHIO (5)
(1)High Temperature Plasma Center, Univ. of Tokyo, (2)Graduate School of Frontier Science, the University of Tokyo,
(3)National Institute for Fusion Science,(4)Research Institute of Superconductivity Kyushu University, (5)Fuji Electric
Systems Co., Ltd.,
An Internal coil device would be expected for exploring high beta plasmas based on plasma relaxation process.
Prof. A Hasegawa proposed an advanced fusion reactor with a dipole configuration, and Mahajan-Yoshida
developed a new high beta state based on two-fluid relaxation theory. To study these high beta plasmas, we
have constructed an internal coil device with a high temperature superconductor. The major radius of the
internal coil is 15 cm, and the coil current is 50 kA. Three different types of Ag-sheathed Bi-2223 tapes are
employed; i.e., a high critical current(Ic=108A at 77K, s.f., 1 micro-V/cm) tape with a low silver ratio for the
main HTS coil, a 0.3wt%Mn-doped Bi-2223 tape for the persistent current switch and 3at%Au-doped Bi-2223
tape for the current lead. The coil is cooled with cold helium gas provided by a GM refrigerator and supplied
to the coil through a check valve. The coil current is directly excited with the external power supply through
removable electrode. It took about 11 hours to cool the coil down to 21K from the room temperature, and the
nominal cable current of 118 A (overall coil current: 50kA) has been achieved. A decay time constant of the
persistent current is a few tens of hours. Weight of the HTS internal coil is 16.8kg. Time constant of motion
for the internal coil is about 70 ms in the center of vacuum vessel at a normal floating position. The position of
the internal coil is monitored with 5 laser sensors which can be detected 5 freedoms (vertical, tilt-X-Y and
sliding-X-Y) of the coil. The resolution of the laser sensor is 10 micrometers. A levitating coil is installed on
top of a vacuum vessel that is made of copper coil. Rating of the levitation coil is 20kA. The vertical position
of the internal coil is feedback-controlled with the regulation of the levitation coil current. The HTS internal
coil is successfully levitated in the vacuum vessel during one hour or more. The accuracy of the internal coil
position is 20 micrometers. Plasma experiments in a dipole configuration have been initiated. The plasma is
produced with 2.45 GHz ECH system. At present, the plasma temperature and density are ~10 eV and
5x1016m-3, respectively.
Corresponding Author:
MORIKAWA JUNJI
morikawa@ppl.k.u-tokyo.ac.jp
Graduate School of Engineering, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku Tokyo 113-8656, Japan
27
- A - Current and Next Step Devices
P3C-A-245
EXPERIMENTAL STUDY OF WATER FLOW DISTRIBUTION INSIDE
TWO-CHANNEL MODEL OF ITER VACUUM VESSEL COOLING
SYSTEM
TANCHUK VICTOR, BABYKIN A., BALUNOV B., CHTCHEGLOV A., GRIGORIEV S., KRYLOV V.
Joint-Stock Company I.I. Polzunov Scientific & Development Association of Research and Design Power Equipment, 3/6 ul.
Atamanskay, St. Petersburg 191167, Russia
VV double wall and neutron shielding plates, poloidal and toroidal ribs with holes for water passing form a
complex system of parallel-series channels for ITER VV cooling. The cooling channels formed in such
manner are characterized by different cross sections (channel heights from 5mm to 90mm), heat loads and
orientation in the gravitational field. At extremely low bulk flow velocities (10?200 mm/s) dimensions and
position of water passage holes in the VV ribs, other VV design and loading conditions could significantly
effect on the water flow distribution inside the parallel channels. This could impact on the VV temperature
state. To investigate the flow distribution in the parallel channels and to prove interchannel flow stability a
two-channel model of a VV test element has been developed. The VV test element is a rectangular box 0.2m
wide, 3m long, with a 2.48 m heated length. The box is divided by an intermediate plate into two channels:
upper 50mm in height and lower 12 mm in height. 3 heaters located at the upper, lower and intermediate walls
produce heat loads separately for each channel. A total of 241 experiments were performed. The obtained
results prove that: (1) design of the channel inlet unit is a decisive factor in water distribution between the
parallel channels at extremely low flow rates; (2) hydraulic friction, inclination angle, water inlet temperature
are not dominant in the mechanism of flow distribution at a significant influence of thermal gravitational
forces; (3) heating of one of the channels has the principal effect on the flow splitting. This effect is especially
drastic at low flow rates (Gtotal ? 0.3-0.4 kg/s), when practically the entire flow comes through a heated
channel (vertical or inclined channels). No reverse circulation has been observed during the tests for all range
of studied flow velocities (10?200 mm/s), so a stable flow distribution is expected for the VV cooling system.
Corresponding Author:
TANCHUK VICTOR
tanchuv@sintez.niiefa.spb.su
The D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA), 3 Doroga na
Metallostroy, Metallostroy, St.Petersburg 196641, Russia
28
- A - Current and Next Step Devices.
P3C-A-262
MAGNUM-PSI, A PLASMA GENERATOR FOR PLASMA-SURFACE
INTERACTION RESEARCH IN ITER-LIKE CONDITIONS
GROOT DE, BART, G.J. VAN ROOIJ(1),V. VEREMIYENKO(1),M.G. VON HELLERMANN(1),C.J.
BARTH(1),G.L. KRUIJTZER(1),J.C. WOLFF(1),H.J.N. VAN ECK(1),W.J. GOEDHEER(1),N.J. LOPES
CARDOZO(1),A.W. KLEYN(1),S. BREZINSEK(2),A. POSPIESZCZYK(2),R.A.H. ENGELN(3)
(1)FOM-Institute for Plasma Physics Rijnhuizen, Assoc. EURATOM-FOM,The Netherlands, www.rijnh.nl(†) (2)IPP, FZ
Jülich GmbH, EURATOM Assoc.,Germany(†) (3)Eindhoven University of Technology,The Netherlands (†)Partners in the
Trilateral Euregio Cluster
Introduction In collaboration with its TEC partners, the FOM-Institute for Plasma Physics is preparing the
construction of Magnum-psi, a magnetized (3 T), steady-state, large area (100 cm^2) high-flux (up to 10^24
H+ ions m^-2s^-1) plasma generator. Magnum-psi is being developed to study plasma-surface interaction in
conditions similar to those in the divertor of ITER and fusion reactors beyond ITER. Magnum-psi will be
embedded in an integrated plasma-surface laboratory including in situ and ex situ, in vacuo surface analysis.
The scientific program includes a strong modeling effort. A pilot experiment (Pilot-psi) has been constructed
to explore the techniques to be applied in Magnum-psi. This contribution addresses the optimization of the
cascaded arc plasma source and the effect of the magnetic field on the expanding plasma beam. Experimental
results achieved on Pilot-psi will be presented to demonstrating that the required hydrogen plasma flux can be
generated with a high-pressure plasma source (cascaded arc) and a longitudinal B-field of 1.6 T. Results of
Pilot-psi: In order to obtain a detailed picture of the plasma fluxes for different cascaded arc plasma source
geometries and magnetic field strengths, we employed electron density measurements by means of Thomson
scattering and the analysis of Stark broadening in atomic emission spectroscopy. Thomson scattering data
yielded radial profiles of the electron density and temperature with a spatial resolution of 1 mm and are in
agreement with high-resolution spectroscopy results. Typical results in hydrogen are: Ne ranging from 10^20
to 1.5*10^21 m^-3 for B=0.4-1.6 T. Te=0.4 eV, only weakly varying with B. Using additional Ohmic heating
of the expanding plasma, the temperature can be increased to a few eV. The flow velocity in the plasma jet
was derived from time of flight analysis of plasma perturbations induced by modulation of the arc current and
found to be subsonic (~250 m/s). Multiplication of the densities and propagation speeds yields hydrogen ion
flux densities well above 10^23 H+ ions m^-2s^-1, proving that even in our pilot experiment ITER relevant
flux densities of hydrogen plasma can be reached, albeit not in steady state (due to the pulsed magnetic field)
and over a small cross section (1 cm^2).
Corresponding Author:
GROOT DE, BART
degroot@rijnh.nl
FOM-Institute for Plasma Physics Rijnhuizen, P.O. Box 1207, 3430 BE Nieuwegein, The Netherlands
29
- A - Current and Next Step Devices
P3C-A-330
THE LASER MÉGAJOULE (LMJ) PROJECT DEDICATED TO
INERTIAL CONFINEMENT FUSION : DEVELOPMENT AND
CONSTRUCTION STATUS.
FLEUROT NOËL, CAVAILLER CLAUDE BOURGADE JEAN-LUC
Commissariat à l'Energie Atomique - Centre DAM Ile de France - BP 12 - 91680 Bruyères le Châtel - France
Selected also for oral presentation
O3B-A-330
The Laser Megajoule (LMJ) facility is a 240 beam facility dedicated to our Inertial Confinement Fusion
program. Its construction was started in 2003 at the French Atomic Energy Commission CESTA center located
near Bordeaux. LMJ is a frequency tripled Nd:glass laser able to focus up to 1.8 MJ – 600 TW of ultraviolet
light (0.35 µm) on targets dedicated to laser matter interaction experiments and to achieve ignition and
ultimately combustion of DT targets in the laboratory. Typical quadruplet focus spot size on target is in the
600 - 700 µm range in diameter and it can be adapted, by using optical phase plates, to obtain elliptical focal
spots. LIL ("Ligne d'Intégration Laser" : the LMJ prototype) has been the first laser in the world to produce 9.5
kJ of UV light in less than 9 ns in 2003 with a single beam. The commissioning of the quadruplet (4 beams) at
0.35 µm is now achieved. We will also present the current LMJ design with its four laser bays (a total of 30
bundles x 8 beams) which produce the infrared light (typically 18 kJ at 1.05 µm per beam at the output of the
amplifier section). Each bundle of 8 beams is then separated in 2 quadruplets in the target bay ; the 60
quadruplets of IR light are frequency tripled at 0.35 µm and focused by large optical gratings through 60 ports
in the 10 m diameter target chamber onto the target. Plasma diagnostics (X-rays, neutrons …) will require
resolutions in the 10 to 100 ps temporal range and 10 µm spatial range to diagnose laser fusion of DT
cryogenic targets in the so called "indirect drive" configuration. LMJ will be able to achieve an energy output
yield of up to 20 MJ. The first contracts concerning both the laser and target chamber area have already been
procured to the French industry.
Corresponding Author:
FLEUROT NOËL
noel.fleurot@cea.fr
Commissariat à l'Energie Atomique - Centre DAM Ile de France - BP 12 - 91680 Bruyères le Châtel - France
30
- A - Current and Next Step Devices
P3C-A-449
JET ENGINEERING: PROGRESS AND PLANS
TODD, THOMAS, KAYE, ALAN PAMELA, JEROME MURARI, ANDREA ROLFE, ALAN RICCARDO, VALERIA
BRENNAN, DAMIAN
EFDA-JET, Culham Science Centre, Abingdon, OX14 3DB, UK
Selected also for oral presentation
O3B-A-449
The UKAEA-Euratom Association has now operated JET for EFDA for over four years, providing a
sophisticated large tokamak facility for experiments run by the Contract of Association institutes. JET
continues to offer a state-of-the-art capability strongly relevant to ITER physics and technology issues,
including beryllium in the torus and a full tritium fuel cycle system. The original 20g inventory of tritium was
re-injected five times for the first DT campaign and 5g of the now remaining 10g was injected and recovered
in the recent Trace Tritium Experiment. Diagnostic and control system developments to track the tritium and
minimise retention in the machine structure continue, with further new systems now being installed. Tritium
operation mandates a remote handling system for work on major in-vessel components while the radiation
field is high, eg 8mSv/hr falling to the “ALARP” target of 350microSv/hr for man entry. JET remote handling
developments continue both in technical aspects such as load transfer control (260kgs at 10m for the poloidal
limiter beams) and in training and rehearsals, ~80% in Virtual Reality since early 2003, and ~20% in the fullscale torus simulator. The VR environment is based on 3D design files, necessitating rigorous design
configuration management for all machine modifications. Facility operation requires machine protection
systems based on sophisticated stress analyses, to constrain operation within boundaries consistent with the
desired plant life. This is especially demanding for Vertical Displacement Events in the high-delta divertor
plasma shapes foreseen for ITER, which generate vertical forces around twice those of the pre-2000 plasmas in
JET. Analyses show that life consumption of the key plant is ~10% at present, with operational limits of 4T,
5MA and 850t VDE force. Collaboration with the Associations has yielded many valuable improvements to
the diagnostic and control systems, eg. real-time control and control of exotic plasma shapes to centimetric
precision. The presently ongoing "Enhanced Performance" shutdown will add a range of capabilities to the
machine including an ITER-like ICRH antenna and improved plasma diagnostic systems. This paper will
detail the principle technical and analytical systems required to meet the challenge of providing an engineering
environment for the JET-EP work programme.
Corresponding Author:
TODD, THOMAS
tnt@jet.uk
Euratom-UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK
31
- A - Current and Next Step Devices.
P3C-A-468
THE JET-ENHANCED PERFORMANCE PROGRAM: MORE HEATING
POWER AND DIAGNOSTIC CAPABILITIES IN PREPARATION FOR
ITER
LIOURE ALAIN, ALAN KAYE (2) ANDREA MURARI (3) JOAQUIN SANCHEZ (4) TOM TODD (2) CARLO
DAMIANI (5) JEROME PAMELA (1)
(1) EFDA JET, Culham, Abingdon, OX14 3EA UK (2) UKAEA JET, Culham, Abingdon OX14 3DB, UK (3) Consorzio RFX,
Corso Stati Uniti, 4, I-35127 Padova Italy (4) CIEMAT, 28040 Madrid, Spain (5) ENEA Brasimone, 40032 Camugnano
(BO), Italy
Since early 2000 the JET-EP program has been aiming at optimising JET for ITER-relevant plasma operations,
from 2005 onwards. The overall heating capability of JET will be increased to 40 MW. The neutral beam
system was up-graded but the major technical challenge is to build an ITER-like ICRH antenna, with
particularly stringent specifications (8 MW/m2 for 10s, compatible with Type-I ELMy H-modes, coupling at
12cm distance to the plasma). The new divertor configuration will be able to absorb more than 300 MJ per
shot. The physics of the power handling will be monitored by sophisticated new diagnostics, e.g. a high-signal
to noise bolometric measurement system and an ambitious IR viewing system using a state of the art camera,
looking at the antenna and the divertor. New halo sensors will be installed to better understand disruption
phenomena. JET will operate in a wider range of plasma conditions. The divertor’s geometry allows hightriangularity ITER-like scenarios (deltaU~0.44, deltaL~0.56) with a greater flexibility with respect to different
plasma configurations. The control of extreme plasma shapes will be re-enforced and a new disruption
mitigation system using a very fast gas valve will be provided. The diagnostic capability will be enhanced by
several new systems designed to address a number of crucial physical phenomena for ITER. To study Tritium
retention further, new technologically challenging erosion-redeposition diagnostics will be installed,
particularly in the divertor region, both real time and integrating. New neutron detectors using the latest
advances in scintillators and data recording techniques will produce much higher count rates and signal to
noise. Detectors for fast á particles with high pitch angle and energy resolution will be installed the closest
ever to the plasma in JET. High-resolution Thomson Scattering, with 20 Hz repetition rate, will provide
temperature and density profiles with a spatial accuracy close to two centimeters. An improved microwave
access will enable broad band reflectometry for density profile and oblique ECE measurement for the first time
on JET. High bandwidth coils and high-n Alfven mode-dedicated diagnostics will allow more emphasis on
MHD regimes. The new diagnostics will be integrated in the JET real-time system. This paper presents an
overview of this program, emphasising the main objectives and pointing out the various technological
challenges and innovations.
Corresponding Author:
LIOURE ALAIN
alain.lioure@jet.efda.org
EFDA, Culham Science Centre, Abingdon, Oxfordshire OX14 3EA (UK
32
- A - Current and Next Step Devices
P3C-A-518
NUCLEAR ANALYSES OF SOME KEY ASPECTS OF THE ITER
DESIGN WITH MONTE CARLO CODES
IIDA HIROMASA, L.PETRIZZI(2) V. KHRIPUNOV(3) G. FEDERICI(1) E. POLUNOVSKIY(1)
(1)ITER Garching Joint Work Site Boltzmannstr. 2 D-85748 Garching Germany (2)Nuclear Fusion Institute, Russian
Research Center "Kurchatov Institute", Moscow, Russia (3)Via E. Fermi 45 00044 Frascati ITALY (Rome)
The design of the ITER machine was presented in 2001 . Radiation transport calculations have been very
important in the assessment of the ITER design, particularly with regard to operational constraints, access for
reactor maintenance and activated waste. A nuclear analysis has been performed on ITER by means of the
most detailed models and the best assessed nuclear data and codes. Calculations have been carried out in a
progression which began with 1D studies for scoping, taking into account the reactor operating conditions,
followed by 2D and 3D calculations taking into account streaming through penetrations, as well as the
complexity of the geometry and the different material thicknesses and compositions. As the construction phase
of ITER is approaching, the design of the main components has been optimsed/finalised and several minor
design changes/optimisations have been made, which required refined calculations to confirm that nuclear
design requirements are met. These have included assessment of nuclear heating in various components during
various phases of the reactor operation, surface heat load on the in-vessel components due to bremsstrahlung
and line radiation from the plasma, nuclear heating and damage of electric insulators due to N-16 in the
blanket and divertor cooling water, and decay gamma-ray dose rate distribution around the machine after
shutdown. This paper reviews some of the most recent neutronic work with emphasis on (i) critical neutronics
responses in the TF coil inboard legs related to design modifications made to the blanket modules and vacuum
vessel; (ii) accurate dose rate calculations after reactor shutdown, to confirm that the shielding around the torus
is sufficient to allow personnel access for machine maintenance. All these detailed Monte Carlo analyses
inevitably require very precise geometry modeling, which demand significant amount of manpower. Some of
the ITER participant teams (in particular, Europe and China) are developing specific tools to facilitate
conversion of CAD drawing information into MCNP models. A brief mention of this activity will be made,
together with anticipated further developments to meet challenges ahead.
Corresponding Author:
IIDA HIROMASA
iidah@itergps.naka.jaeri.go.jp
ITER Naka Joint Work Site,c/o JAERI,Naka-machi, Naka-gun,Ibaraki-ken,Japan
33
- A - Current and Next Step Devices
P3C-A-522
TRANSPORT, LOGISTICS AND PACKAGING OF ITER COMPONENTS
GUÉRIN OLIVIER, B. COUTURIER (1) A. MAAS (1) AND EISS TEAM
(1) Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France)
The construction of ITER will be an important challenge over the coming years. Components for the machine
will be manufactured by all ITER partners, in factories around the world. These components, some of them
very large and heavy, will have to be transported to the ITER construction site. In the case of the European site
for ITER, at Cadarache in the South-East of France, the transport will have to be ensured over an itinerary of
around 100 km, from the nearest industrial harbour to the site. Extensive studies have been undertaken in
various fields, including the choice of an itinerary and its optimisation, the use of barges, ships, trucks, trailers
and handling tools, kinematics and logistics of transports, packaging of different ITER components. Detailed
logistics studies have been performed with world-leading companies in this field. An important feedback from
a similar technical challenge, the successful completion in time and budget of an itinerary between Bordeaux
and Toulouse for the transport of the future Airbus A380 parts, has also been used. The feasibility of these
transports has been demonstrated and the different aspects of the studies retained solutions will be described in
the paper.
Corresponding Author:
GUÉRIN OLIVIER
olivier.guerin@cea.fr
Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France
34
- A - Current and Next Step Devices
P3C-A-523
STUDIES FOR SITE PREPARATION FOR ITER CONSTRUCTION
FARDEAU AGNES, F. BLANC (1) J.-D. CARDETTINI (1) J.-R. MANDINE (1) R. GUÉRIN (1) L. PATISSON (1)
P. BERGÉGÈRE (1) A. SANTAGIUSTINA (2) P. GARIN (2) AND EISS TEAM*
(1) Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France (2) Association EuratomCEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France
The implantation of a nuclear facility as ITER (surface of 40 hectares) requires many preparatory studies and
works, particularly with respect to: Underground characterisation (geological survey) Impact of seismic
hazard on design Topography, layout Climate data (mainly for the design of buildings and systems)
Deforestation, excavations Networks, fences and roads Definition of an area on or close to the site for the
companies during the construction phase The aim of this paper is to present the main results of the studies and
works carried out within the European ITER Site Studies framework. To perform these studies, all needs of
ITER have been taken into account (ITER requirements and design assumptions), but the proximity of the
CEA centre has also been valorised. To choose the site for ITER implantation, detailed geological and
geophysical investigations have been carried out (60 drillings, 4 km of seismic refraction lines, several tests on
samples). Then, taking into account the meteorological data available since 1960 (particularly the main wind
direction) and the topography (based on an aerial photos and topographic surveys), buildings and roads have
been implemented, on 4 platforms (in order to minimize excavation work). Similarly, detailed studies have
been carried out to implement all and satisfy ITER needs in terms of: cooling water supply (6,700 m3/day),
potable water supply (400 m3/day), sanitary sewage (200 m3/day), industrial sewage (200 m3/day), cooling
sewage (blow down: 3000 m3/day) treatment and exhaust, rainfall network, electrical supply (120 MW of
continuous electrical power). Concerning the above items, existing infrastructures of CEA centre could be
used, leading to substantial savings. Finally, ITER buildings, as defined in the generic site, have been
estimated insufficient, with regard to the character of the project and other buildings, offered by Europe to the
ITER partners, have been identified for the construction phase, but also for the exploitation phase. Preliminary
studies have been carried out to define: a Welcome Centre (for visitors or workers families), a restaurant, a
medical building, and an access control building.
Corresponding Author:
FARDEAU AGNES
fardeau@dircad.cea.fr
Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France
35
- A - Current and Next Step Devices
P3C-A-531
READINESS OF CADARACHE FOR STARTING ITER CONSTRUCTION
LYRAUD CHARLES, J.-M. BOTTEREAU (1) A. FARDEAU (2) O. GUÉRIN (1) A. MAAS (1) S. MATTEI (2) P.
GARIN (1) AND EISS TEAM
(1) Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France (2) Direction de l’Énergie
Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France
Since the beginning of the European ITER Site Studies in 2001, specific attention has been paid to the
readiness and preparedness of the European site. These aspects include technical preparation of the site and its
surrounding, as well as the welcome of the first international team members and their families in Provence in
the best possible conditions. The purpose of this paper is to present the work already carried out and to be
performed to ensure the successful construction of ITER in Europe, within time and budget. This will cover
technical and socioeconomic aspects, such as: the licensing process, the increase of the industrial
environment awareness, the heavy load itinerary road modification programme, the site preparation, the
annex buildings offered by the host, the large poloidal field coil manufacturing facilities, erected on site to
minimise the manufacturing and handling risks, the international school development (Japanese, Chinese,
Russian, Korean and European languages) the housing for several hundred foreign families, the set up of a
communication and welcome organisation. Phase 1: Starting on the site decision date, an ITER temporary
facility to welcome ITER team members will be set up on the Cadarache site, where all services are already
available for 5,000 people. Until the creation of the ITER Legal Entity and the European Legal Entity, the
temporary International Team will complete the ITER construction filesThe European team will deal with
public enquiries, site works, deforestation and site levelling, annex buildings final studies and start of works,
enterprise yard for construction on site, heavy load itinerary works, industrial environment awareness of the
thousands of companies located around the site. The same activity will be performed at the European level and
World level by the Partners. A close follow-up will be carried out to supervise the availability of the
International School soon after ILE creation in close collaboration with educational authorities of the ITER
Parties. The licensing process will lead to the authorisation of construction of ITER facilities (French
governmental decree). Phase 2: Starting at ILE creation. The annex buildings (Welcome Centre, restaurant,
first aid facilities, offices…) as soon as they become available will be offered to the ITER organisation, as they
are not linked to the licensing process. ITER building construction programme can be launched.
Corresponding Author:
LYRAUD CHARLES
charles.lyraud@cea.fr
Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France
36
- B - PLASMA HEATING AND CURRENT DRIVE.
P3T-B-19
THE ALCATOR C-MOD LOWER HYBRID CURRENT DRIVE
EXPERIMENT TRANSMITTER
MONTGOMERY GRIMES, DAVID TERRY RON PARKER DEXTER BEALS
Same as corresponding address
Alcator C-Mod, is a high-field, high-density, diverted, compact tokamak, which, in its present form uses
inductive current drive and is heated with 5 MW of ICRF auxiliary power. C-Mod is in the process of being
upgraded with a 4.6 GHz Lower Hybrid heating and current drive system. The purpose of the experiment is to
develop and explore the potential of “Advanced Tokamak Regimes” under quasi-steady-state conditions. In
this paper, an overview of the RF transmitter and the controls and protection systems for the Lower Hybrid
Project is given. The transmitter will use twelve 250 kW klystrons operating simultaneously which will result
in a total nominal power at the klystrons of nearly 3 MW for a planned pulse width of 5 seconds. Active
control system vector modulators provide phase and amplitude drive for each klystron, and I-Q detectors are
used to monitor phase and amplitude. These feedback signals are used in digital controllers for closed-loop
control of klystron phase and amplitude to preset values. An expected upgrade of four additional klystrons will
result in a total nominal power of 4 MW. The transmitters have been tested to full power, and installation of
the Lower Hybrid Current Drive experiment on the C-Mod Tokamak is expected in 2004.
Corresponding Author:
MONTGOMERY GRIMES
grimes@psfc.mit.edu
MIT Plasma Science and Fusion Center, 190 Albany St., Cambridge, MA 02139 USA
37
- B - Plasma Heating and Current Drive.
P3T-B-25
DESIGN OF AN ULTRA-BROADBAND SINGLE-DISK OUTPUT
WINDOW FOR A FREQUENCY STEP-TUNABLE 1 MW GYROTRON
XIAOKANG, YANG, GUENTER DAMMERTZ (1A) ROLAND HEIDINGER (1B) KAI KOPPENBURG (1A) FRITZ
LEUTERER (3) BERNHARD PIOSCZYK(1A) DIETMAR WAGNER (3) MANFRED THUMM (1A),(2)
(1) Forschungszentrum Karlsruhe, Association EURATOM-FZK, (a) IHM, (b) IMF-1 76021 Karlsruhe, Germany. (2)
Universitaet Karlsruhe, IHE, 76128 Karlsruhe, Germany. (3) Max-Planck-Institut fuer Plasmaphysik, Association
EURATOM-IPP,85748 Garching, Germany
For plasma stabilization in the ASDEX-Upgrade tokamak, there is interest in step-tunable gyrotrons operating
at frequencies between 105 GHz and 140 GHz. For this purpose a multifrequency gyrotron is under
construction at Forschungszentrum Karlsruhe (FZK) in a cooperative parallel development with the Institute of
Applied Physics in Nizhny Novgorod, Russia. Output window design is one of the key issues to realize
broadband output of a multi-frequency gyrotron. Corresponding to the development of such frequency steptunable 1 MW gyrotrons at FZK, this paper summaries recent development of broadband single-disk output
windows, in particular the Brewster window with a CVD-diamond disk. The thickness of the disk has to be
optimized to get low power reflection over a broadband incident angle range around the Brewster angle.
Detailed calculations of the transmission characteristics for the CVD-diamond disk Brewster window have
been performed for the all considerd 9 modes from TE17,6 at 105 GHz up to TE23,8 at 143 GHz, and for
thickness of the disk from 1.5 mm up to 2.0 mm. Calculations show that it is difficult to choose the disk
thickness of a CVD-diamond Brewster window for this frequency step-tunable gyrotron, since the choice
depends on both the most important frequencies and the availability of the disks. If one prefers to place the low
reflection area in the middle of the discussed frequency range, such as 120-130 GHz, the thickness of 1.6 mm
is near optimum and its -20 dB bandwidth angle is more than 30 degrees. For operation near 105 GHz and 140
GHz, a 1.9 mm disk is preferable. Its -20 dB bandwidth angle is around 30 degrees, but for other central
frequencies, the situation is not so good. Further calculation results also show that the -20 dB bandwidth angle
decreases with increasing disk thickness from 1.5 mm to 2.0 mm. However, thin CVD-diamond disks will add
mechanical problems to the window construction. Another important factor to be considered is the analysis of
the bow and maximum tensile stresses in brazed windows arising from differential pressure uniformly applied
over the surface of the disks, when they are not operated in evacuated transmission systems.
Corresponding Author:
XIAOKANG, YANG
xiaokang.yang@ihm.fzk.de
Forschungszentrum Karlsruhe, Association EURATOM-FZK, IHM, D-76021 Karlsruhe, Germany
38
- B - Plasma Heating and Current Drive.
P3T-B-51
EXPERIMENTS ON A 170 GHZ COAXIAL CAVITY GYROTRON
PIOSCZYK, BERNHARD, ANDREAS ARNOLD (2), HERBERT BUDIG (1A), GUENTER DAMMERTZ (1A),
OLGIERD DUMBRAJS (3), ROLAND HEIDINGER (1B), STEFAN ILLY (1A), JIAMBO JIN(1A), GEORG MICHEL
(4), TOMASZ RZESNICKI (1A), MANFRED THUMM (1A,2), XIAOKANG YANG (1A)
(1a,b)FZK Karlsruhe, (a) IHM, (b) (IMF I), D-76021 Karlsruhe, Germany (2)Universitaet Karlsruhe, IHE, D-76128
Karlsruhe, Germany (3) Helsinki University of Technology, FIN-02150 Espoo, Finland (4) MPI fuer Plasmaphysik, D-17491
Greifswald, Germany
Within a development program performed as an ITER task at the Forschungszentrum Karlsruhe (FZK) the
feasibility of manufacturing a multi-megawatt coaxial gyrotron operated in continuous wave (CW) has been
investigated and information necessary for a technical design and industrial manufacturing has been obtained.
Based on these results the development of a coaxial cavity gyrotron with an RF output power of 2 MW, CW at
170 GHz as could be used for ITER is in progress in cooperation between EURATOM Associations (CRPP
Lausanne, FZK Karlsruhe and HUT Helsinki) together with European tube industry (Thales Electron Devices,
Velizy, France). In parallel to that work on a first industrial prototype tube, the previously used short pulse 165
GHz, TE31,17 coaxial cavity gyrotron at FZK has been modified for operation at 170 GHz in the TE34,19
cavity mode. The modified experimental gyrotron operates in the same mode as foreseen for the industrial
prototype and uses a cavity with same dimensions. In addition, the gyrotron is equipped with an improved
quasi-optical RF output system same as designed for the prototype. The experimental operation is planned to
start within the next weeks. The investigations have two main goals: (1) to verify experimentally the design of
the main components of the industrial prototype by studying both the efficiency of RF generation and mode
competition and the properties of the quasi-optical RF output system, (2) to provide a high power, short pulse
(~5-10 ms) test possibility for studying a prototype of the remotely steerable launcher of the upper ITER port
plug for neoclassical tearing mode stabilization. Results concerning as well the gyrotron operation and the
conditions for the launcher test are expected and will be reported.
Corresponding Author:
PIOSCZYK, BERNHARD
bernhard.piosczyk@ihm.fzk.de
Forschungszentrum Karlsruhe, Association EURATOM-FZK, D-76021 Karlsruhe, Germany
39
- B - Plasma Heating and Current Drive.
P3T-B-78
THE UPGRADE OF THE DIII-D EC SYSTEM USING 120 GHZ ITER
GYROTRONS
CALLIS, R.W., J. LOHR (1), Y.A. GORELOV (1), D. PONCE (1), K. KAJIWARA (2), AND J.F. TOOKER (1)
(1) General Atomics, P.O. Box 85608, San Diego, California, 92186-5608 (2) Oak Ridge Institute for Science Education, Oak
Ridge, Tennessee
The planned growth in the EC system on DIII-D over the next few years requires the installation of two
depressed collector gyrotrons, a high voltage power supply, two low loss transmission lines, and the required
support equipment. Although the original system is based on a frequency of 110 GHz, there is a benefit to the
US Gyrotron development program, and the US ITER EC hardware manufacturer, if the next generation of EC
equipment for the DIII-D program adopts the 120 GHz ITER startup frequency. This new DIII-D EC
equipment could then be considered as a prototype of the ITER EC Startup System. By building the DIII-D
hardware to the ITER specifications it would allow the US ITER program to gain beneficial prototyping
experience on a working tokamak, prior to committing to building the hardware for delivery to ITER. *Work
was supported by the U.S. Department of Energy under DE-FC02-04ER54698 and DE-AC05-76OR00033.
Corresponding Author:
CALLIS, R.W.
callis@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
40
- B - Plasma Heating and Current Drive.
P3T-B-82
THE LHCD LAUNCHER FOR ALCATOR C-MOD – DESIGN,
CONSTRUCTION, CALIBRATION AND TESTING*
J. HOSEA (1), W. BECK (2) S. BERNABEI (1) R. CHILDS (2) R. ELLIS (1) E. FREDD (1) N. GREENOUGH (1)
M. GRIMES (2) D. GWINN (2) J. IRBY (2) P. KOERT (2) C. C. KUNG (1) G. D. LOESSER (1) R. PARKER (2)
D. TERRY (2) R. VIEIRA (2) J. R. WILSON (1) J. ZAKS (2)
(1) Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ, USA (2) Plasma Science and Fusion Center,
MIT, Cambridge, MA, USA
MIT and PPPL have joined together to fabricate a high power lower hybrid current drive (LHCD) system for
the Alcator C-MOD device to help support quasi steady-state AT regimes. A 3 MW source and a single
launcher system have been provided for initial experiments. The launcher consists of a 24-column by 4-row
waveguide array and has independent phasing control for each of the columns to maximize spectral control [1].
It was designed and constructed to support the application of 1.5 MW for up to 5 sec to the plasma, based on
previous experimental power limits, and possibly 2 MW with sufficient conditioning. Some of the launcher
design was based on previous experience with other devices: e.g., brazing of alumina windows into titanium
guides is used to provide isolation of the coupler arrays at the plasma from the power feed guide system -thereby facilitating the spectral control for the power launched into the plasma. However, much of the design
uses new concepts for maximizing the number of guides in the relatively narrow C-MOD port while also
maximizing the total power handling capability. Stacked waveguides incorporating a two-hole sidewall splitter
design are used to deliver the power to the couplers [2]. All gaskets (microwave seals) are located outside the
vacuum, and the alumina windows are “tuned” to the system frequency of 4.6 GHz [3]. Construction,
calibration and testing techniques and results used in the carrying out of the design will be discussed. In
particular, the bolt/gasket design for attaching the coupler to the stacked waveguide, the brazing of the alumina
windows into the titanium couplers, and the power splitter design required considerable analysis and
prototyping to achieve the desired performance. In addition, the results of high power tests for each of the
component sections of the launcher assembly will be presented. These tests have been successfully conducted
to power levels (in the range of 100 kW) representative of the maximum voltage/current conditions that will be
experienced on C-MOD. *Work supported by US DOE Contracts No. DE-AC02-76CH03073 and DE-FC0299ER54512 1. S. Bernabei et al., Fusion Science and Tech., 43, 145 (2003) 2. C. Kung et al., Proceedings of
the 20th IEEE/NPSS SOFE Conf., P3-21 (San Diego, 2003) 3. J.R. Wilson et al., 15th Top. Conf. On RF
Power in Plasmas, AIP Proc. Vol. 694, 283 (2003)
Corresponding Author:
J. HOSEA (1)
jhosea@pppl.gov
Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ, USA
41
- B - Plasma Heating and Current Drive.
P3T-B-94
DESIGN AND OPERATION OF THE WENDELSTEIN 7-X ECRH HIGH
VOLTAGE POWER SUPPLIES
JÜRGEN ALEX, MICHAEL BADER (1) HARALD BRAUNE (2) DR. VOLKER ERCKMANN (2) RÜDIGER
KRAMPITZ (2) GEORG MICHEL (2) MARC MÜLLER (1) FRANK NOKE (2) DR. GÜNTER PFEIFFER (2)
FRANK PURPS (2) EDGAR SACHS (3) MARIO WINKLER (2)
(1) Thales Broadcast & Multimedia, Bahnhofstr. 34, 5300 Turgi, Switzerland (2) Max-Planck Institut für Plasmaphysik
(IPP), Wendelsteinstr. 1, 17491 Greifswald, Germany (3) FEAG, A Siemens Company, Günther-Scharowsky-Str. 2, 91058
Erlangen, Germany
The high voltage power supplies for the heating systems of Wendelstein 7-X are universal systems to be used
on either ECRH or NBI heating. All power supplies are connected to a switching system, allowing to supply
any load from any power supply. The power supplies are of the pulse-step-modulator type and rated for up to
130 kV / 130 A. The complete system has been delivered by a consortium between Thales Broadcast &
Multimedia and Siemens. The tests on the first system were finished in November 2003. Since then the first
power supply has been in operation for the tests on the first gyrotron on site. The paper gives an overview on
the results of the power supply testing and the operation on the gyrotron. It shows the performance under
normal operation as well as the short-circuit switching-off behaviour.
Corresponding Author:
JÜRGEN ALEX
juergen.alex@thales-bm.ch
Thales Broadcast & Multimedia, Bahnhofstr. 34, 5300 Turgi, Switzerland
42
- B - Plasma Heating and Current Drive.
P3T-B-113
THERMAL ANALYSIS AND OHMIC LOSS ESTIMATION OF
POLARIZER FOR ITER ECCD SYSTEM
SAIGUSA MIKIO, K. TAKAHASHI(2), Y. KASHIWA(1), S. OISHI(1), Y. HOSHI(1), T. NAKANO(1), A.
KASUGAI(2), K.SAKAMOTO(2), T. IMAI(2)
(1)Ibaraki University, Nakanarusawa 4-12-1, Hitachi-shi, Ibaraki-ken, Japan, (2)Japan Atomic Energy Research Institute,
Naka-machi, Naka-gun, Ibaraki-ken, Japan.
An electron cyclotron current driving (ECCD) method is useful for suppressing the neoclassical tearing modes
which degrade the energy confinement of tokamak plasmas. ECCD system in International Thermonuclear
Experimental Reactor (ITER) needs to optimize polarization for exciting pure ordinary wave at an oblique
injection into the tokamak plasmas. The specification of ECCD system in ITER demand severe operational
conditions for transmission lines and polarizers, that is 1MW per one wave guide. Therefore it is important to
evaluate ohmic loss of the rectangular grooved mirror installed in a miter bend type polarizer. The several
polarizers were made of chromium copper alloy, installed in miter bends and tested at 170 GHz, 441kW during
6 seconds. The increase in temperature on the back plate of the grooved mirror has been measured with thermo
couplers. The predicted dependences of ohmic loss of grooved mirrors on mirror rotation angle and the
rotation angle of the polarization plane of the incident waves agree with the experimental results, qualitatively.
The thermal analysis of grooved mirror has been performed with the 3D FEM code: FEVA, so that the
behavior of the grooved mirror temperature could be explained.
Corresponding Author:
SAIGUSA MIKIO
saigusa@ee.ibaraki.ac.jp
Ibaraki University, Nakanarusawa 4-12-1, Hitachi-shi, Ibaraki-ken, Japan
43
- B - Plasma Heating and Current Drive.
P3T-B-152
TESTS OF LOAD-TOLERANT EXTERNAL CONJUGATE-T
MATCHING SYSTEM FOR A2 ICRF ANTENNA AT JET
IGOR MONAKHOV, A.WALDEN (1) T.BLACKMAN (1) D.CHILD (1) M.GRAHAM (1) W.HARDIMAN (1)
P.U.LAMALLE(2) M.NIGHTINGALE (1) A.WHITEHURST (1) JET EFDA CONTRIBUTORS (3)
(1) Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK (2) LPP-EPM/KMS,
Association Euratom-Belgian State, Brussels, B-1000, Belgium (3) Appendix of J.Pamela, et al., Fusion Energy 2002, IAEA,
Vienna (2002)
Antenna matching during strong and fast loading perturbations introduced by ELMs is one of the major
challenges of high power ICRF operations in H-mode plasmas both on present-day tokamaks and on ITER.
The principle of conjugate-T matching offers a promising general approach to the problem and the methods for
its implementation on tokamaks have been the focus of attention lately. A new 'ITER-like' antenna based on
in-vessel conjugate-T matching by vacuum capacitors is being developed on JET [1]. A complementary
technique to improve the ELM-tolerance of the existing JET A2 antennas by using the external conjugate-T
circuit tuned by coaxial phase-shifters ('trombones') was proposed recently [2]. The latter approach relies
entirely on well-established coaxial line technology; it also opens an opportunity to conjugate the straps
belonging to different antenna arrays and, thus, to ensure an arbitrary array phasing. An upgrade of RF plant
incorporating the external conjugate-T matching of two A2 four-strap antenna arrays into the existing JET RF
system is now under consideration. In order to make a 'proof-of-principle' assessment of the proposal in
realistic conditions a prototype system was installed and tested at JET. The set-up involved one pair of
adjacent straps of the same antenna array powered by a single RF amplifier. The experimental program
consisted of network analyser circuit characterisation, high voltage tests in vacuum and plasma operations in a
range of scenarios including Type I ELMy H-mode. The tests confirmed the feasibility of the proposed
matching scheme both for vacuum and plasma loading. Clear indications of high load tolerance during
sawteeth and ELMs were observed in agreement with circuit simulations. Reliable trip-free performance was
demonstrated in the 32-51 MHz frequency band at <1MW power levels. The paper provides a summary of the
recent external conjugate-T matching activities at JET with an emphasis on the prototype test results. The work
was performed under the European Fusion Development Agreement, and jointly funded by the UK
Engineering and Physical Sciences Council and by EURATOM. [1] F. Durodie, et al., Proc. 15th Top. Conf.
RF Power in Plasmas, Moran, 2003, AIP 694, 98 [2] I. Monakhov, et al., Proc. 15th Top. Conf. RF Power in
Plasmas, Moran, 2003, AIP 694, 150
Corresponding Author:
IGOR MONAKHOV
igor.monakhov@jet.uk
Euratom/UKAEA Fusion Association, J20/1/3, Culham Science Centre, Abingdon, OX14 3DB, UK
44
- B - Plasma Heating and Current Drive.
P3T-B-156
NEUTRONIC ANALYSIS OF ITER NEUTRAL BEAM TEST BED
MICHAEL LOUGHLIN,
It is proposed that ITER have at least two (and possibly three) heating neutral beam injectors. These will inject
1MeV deuterons in to the plasma and are expected to operate for periods of up to one hour. This represents a
major technological step forward. It is therefore necessary to operate a test during the construction phase of
ITER so that a working reliable system is available for the early operational phase. During the testing of the
injector, deuterons will be fired in a calorimeter. This will result in the build up of deuterium and then the
production of 2.5MeV neutrons via the D(d,n)3He reaction, A second branch of the d-d reaction (D(d,p)T)
produces tritium and then the reaction D(t,n)4He produces 14MeV neutrons. Calculations indicate that the
14MeV neutron production is less than 1% of the total neutron yield. The total neutron production of the test
bed facility is estimated to be 1022 neutrons. Neutron transport calculations are therefore important for the
determination of activation of machine structures, dose to workers during maintenance, and the design of
shielding around the device. This paper describes the results of these calculations. The neutron transport was
modelled using the Monte-Carlo particle transport code MCNP. This was used to determine the neutron fluxes
and spectra throughout the injector and in ancillary apparatus around it. The activation of the components was
calculated using the inventory code FISPACT. Dose estimates were then made using further gamma transport
calculations, again using MCNP. It is found that substantial shielding will be needed around the device and no
access will be possible within this area during operations. At the end of operations some components will be
sufficiently activated that special work practices will be required to allow maintenance while keeping the dose
to workers below regulatory limits. It is recommended that steps be taken to minimise the build up of
deuterium within the calorimeter to reduce the neutron production and that the use of low activation steels be
considered for some items close to the neutron source. The 14MeV neutrons, although produced at a lower
level, are shown to produce significant additional activation via threshold reactions. This work was funded
jointly by the United Kingdom Engineering and Physical Sciences Research Council and by Euratom.
Corresponding Author:
MICHAEL LOUGHLIN
michael.loughlin@ukaea.org.uk
UKAEA Fusion, Culham Science Centre, Abingdon, Oxfordshire, OX14 1PR, UK
45
- B - Plasma Heating and Current Drive.
P3T-B-160
A REVIEW OF JET NEUTRAL BEAM SYSTEM PERFORMANCE 1994
TO 2003
ROBERT KING, CLIVE CHALLIS, DRAGOSLAV CIRIC
UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK
The operational performance of the JET Neutral Beam Injector (NBI) system during 2003 is presented and
compared with NBI operation from 1994 to 2002. The paper also addresses different demands imposed on NBI
operation during the JET Joint Undertaking (until the end of 1999) and the European Fusion Development
Agreement (EFDA) JET Operating Contract (from 2000). The JET experimental programme in 2003 consisted
of six experimental campaigns including high power, trace tritium and reverse field. The NBI system was used
either for auxiliary plasma heating, or in support of various plasma diagnostics, on each of the 141 campaign
days. In addition, the NBI system was operated for a further 68 commissioning days. Octant 4 Neutral Injector
Box (NIB4) was operated using six 80kV/52A Positive Ion Neutral Injectors (PINIs), one 130kV/60A PINI
and one 140kV/30A PINI. Octant 8 NIB was equipped with eight 130kV/60A PINIs, but due to installation
and commissioning of two new 130kV/130A power supplies, only four were available before August 2003.
Two more were brought into operation in August 2003 and the remaining two in November 2003. The
performance figures for 1994 to 2001 were achieved with 16 PINIs. The material presented in the paper shows
new operational performance records achieved in 2003, derived from data focused on average and maximum
pulse lengths pulse power and injected pulse energy. During the 2003 JET experimental programme the NBI
system was used to inject energy into ~2900 JET plasma pulses. The total energy injected was 163GJ, with
total beam injection time in excess of 19500s. Over the last ten years the issue of JET NBI PINI reliability and
availability has been of significant interest. A discussion is presented where terminology is defined, specific
technical systems causing unreliability and non-availability are analysed and operational practices are
reviewed. The performance analysis shows that during the period of JET operation under the EFDA contract,
the NBI facility has successfully changed from high power - short pulse to high power - long pulse (10s)
operation. It also shows that the sources of unreliability and non-availability have largely remained constant
during this change. In particular, it is noted that the new Power Supplies have very rapidly achieved reliable
operation. Conclusions are drawn on the importance of structured Commissioning procedures. The work is
funded by EURATOM through the EFDA JET Operating Contract.
Corresponding Author:
ROBERT KING
robert.king@ukaea.org.uk
UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK
46
- B - Plasma Heating and Current Drive.
P3T-B-165
DEVELOPING A FULL SCALE ECRH MM-WAVE LAUNCHING
SYSTEM MOCK-UP FOR ITER
ELZENDOORN BARTHOLOMEUS QUIRINUS SEBASTIANES, M.P.A. VAN ASSELEN (1), W.A. BONGERS (1),
J.W. GENUIT (1), M.F. GRASWINCKEL (1), R. HEIDINGER (2), B. PIOSCZYK (2), T.C. PLOMP (1), D.M.S.
RONDEN (1), A.G.A. VERHOEVEN (1).
(1) FOM Rijnhuizen (2) Forchungs Zentrum Karlsruhe
An ECRH (electron-cyclotron resonance heating) launching system for the ITER upper ports is being
designed. The aim of the system is to inject Electron Cyclotron Waves (ECW) in the ITER plasma in order to
stabilize neoclassical tearing modes (NTM). Each upper-port launcher consists of eight mm-wave lines each
capable of transmitting high power up to 2 MW at 170 GHz. To avoid movable mirrors at the plasma-facing
end of the launcher, the concept of remote mm-wave beam steering (RS) is used. The mock-up consists of a
full-scale mm-wave system placed in a vacuum environment. The mock-up foresees in two separate vacuum
systems, which simulates primary or torus vacuum and secondary vacuum. Secondary vacuum is required for
the partly quasi-optical mm-wave beam trajectory and to provide a second tritium boundary. A diamond
window will provide the first tritium confinement. A phased testing plan is made in order to end in Lausanne
in 2006 for CW full power tests. CW high power tests require cooling on each mm-wave component. The
mock-up requires two separate cooling systems. The cooling system for the square corrugated and the fixed
mirror will also be used to simulate ITER baking conditions with a coolant temperature of 240 ºC and with a
pressure of 4.4 MPa. The second cooling system provides cooling for the mm-wave components placed in
secondary vacuum. The operational temperature for the transmission line is 150 ºC the estimated coolant
pressure 1 MPa. The operation temperature in the secondary vacuum containment at ITER is 100 ºC, this
temperature will be provided by heating blankets. The mm-wave system will be tested under full power CW
operation; these tests will provide information about surface temperatures of mirrors and wall thermal loading
of the square corrugated waveguide. The systems efficiency will be established by calorimetric measurements,
and measuring antenna patterns. The lifecycle tests under influence of temperature variations, realistic coolant
pressures and in a vacuum atmosphere will give the last information before the detailed final design of the
ECRH launching systems can start. ‘This work, supported by the European Communities under the contract of
Association between EURATOM/FOM, was carried out within the framework of the European Fusion
Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the
European Commission.’
Corresponding Author:
ELZENDOORN BARTHOLOMEUS QUIRINUS SEBASTIANES
ben@rijnh.nl
FOM Rijnhuizen, Edisonbaan 14, 3430 BE, Nieuwegein, The Netherlands.
47
- B - Plasma Heating and Current Drive.
P3T-B-171
DIGITAL MOCK-UP DESIGN OF THE REMOTE STEERABLE ITER
ECRH LAUNCHING SYSTEM
D.M.S. RONDEN, W.A. BONGERS (A), A. BRUSCHI (C), I. DANILOV (B), B.S.Q. ELZENDOORN (A), J.W.
GENUIT (A), M.F. GRASWINCKEL (A), G. HAILFINGER (B), R. HEIDINGER (B),T.C. PLOMP (A) AND A.G.A.
VERHOEVEN (A)
(a) FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Edisonbaan 14, 3439MN Nieuwegein, The
Netherlands (b) Forschungszentrum Karlsruhe GmbH (c) CNR Institute for Plasma Physics, Milan
The design of a digital mock-up of the remote steerable ECRH (Electron Cyclotron Resonance Heating) top
launcher is a vital part of the entire development process when working in a complex environment such as
ITER. The aim is to have a digital model available that at all times represents the latest developments of the
overall design. The ECRH top launcher will consist of up to 8 beam lines per upper port, each capable of
delivering up to 2 MW, 170 GHz of ECW (Electron Cyclotron Wave) power to the plasma, primarily for the
stabilization of NTM’s (Neoclassical Tearing Modes). For this, the need arises for the beams to be steerable.
The steerable mirror mechanism is considered a critical component for the ECH&CD system since it is subject
to ECW, neutron, magnetic and thermal loads and hence must be cooled during plasma operation. To avoid the
need of placing the mirror’s steering mechanisms at the plasma-facing end of the launcher, the concept of
remote mm-wave beam steering (RS) is used, having a corrugated square waveguide within the port-plug and
the steerable optic is then placed outside of the first confinement boundary (provided by water-cooled diamond
windows) of the vacuum vessel. Also a fixed mirror is placed at the end of each waveguide - inside the front
shielding blanket - to steer the beam in the right direction. Careful placement of these mirrors is essential to
limit the size of the shielding blanket penetration for the ECW-beams to pass through, which has to be kept to
an absolute minimum. Since many of these design parameters are still converging to an optimum, a conceptual
3d-model has been created that requires little time to update. This has been accomplished to such a high degree
that individual parameter values derived from the physics of millimetre wave beam propagation can be
modified inside the model and will result in a direct visual feedback on the dimensional impacts that
modifications have on the launcher’s total structure. Another accomplishment of the digital mock-up has been
its capability to make highly accurate and adaptive beam tracings and to create complex curved mirror
surfaces. This work is being carried out under the EFDA technology research programme activities, EFDA
technology task TW3-TPHE-ECHULA and B1, with financial support from NWO.
Corresponding Author:
D.M.S. RONDEN
ronden@rijnh.nl
FOM Institute of Plasma physics Rijnhuizen, P.O.Box 1207, 3430BE Nieuwegein, The Netherlands
48
- B - Plasma Heating and Current Drive.
P3T-B-172
PRE STUDY RESULTS ON HIGH VOLTAGE SOLID-STATE SWITCHES
FOR GYROTRON PROTECTION.
A.B. STERK, A.G.A. VERHOEVEN(1), T. BONICELLI(2), D.FASEL(3), A.WELLEMAN(4), S. GEKENIDIS(4)
(1)FOM institute, Nieuwegein, The Netherlands (2)EFDA-CSU, Max Planck Institute, Garching, Germany. (3)CRPP,EPFL,
Lausanne, Switzerland. (4)ABB Switzerland Ltd Semiconductors, Lenzburg, Switzerland.
Introduction A pre study to develop a concept for a high voltage semiconductor switch as a protection for
gyrotrons or klystrons has been launched by the FOM institute in close co-operation with the industrial partner,
ABB semiconductors, Switzerland. Solid-state switches are included in the latest ITER reference designs both
for the Electron Cyclotron Heating and Current Drive and the Lower Hybrid H&CD systems. The first
(prototype) switch is specified for the European Gyrotron test stand that will be built at CRPP in Lausanne.
Description and Design parameters The main function of the solid-state switch is to protect the gyrotron in
case of an electric arc in the cavity by disconnecting the gyrotron from the main power supply. The operating
time must be no longer than 10 microsec. The total energy deposit by the arc in the gyrotron must be limited to
10 Joule. An additional function is to achieve modulation of the gyrotron power by modulating the main power
supply voltage. On-Off modulation frequencies up to 5 kHz are possible in CW operation. Design parameters
Technology Solid State (IGBT) Rated load voltage 60 kVDC Isolation voltage to ground 120 kVDC (10 min.)
Rated load current 80 A Trip current level < 100 A Peak current limitation < 1 kA Recovery time after shortcircuit < 200 ms Fast switch-off time < 10 microsec. Modulation frequency 5 kHz Current wave form Square
wave, duty cycle range: 10-90% ON From the two candidates considered, IGBT and IGCT, the IGBT
technology is chosen as the most favourable because of their much lower switching losses at high repetition
rate (5 kHz) and the controllability through the gate at low power levels. For the IGBT solution two voltage
levels are investigated (2.5 and 5.2 kV). Based on the simulation results the 2.5 kV press-pack IGBT is chosen.
A 10 kV prototype assembly has been build and extensive measurements are executed. The switch-on and
switch-off time, the voltage distribution and the behavior under short-circuit conditions are investigated. All
possible fault conditions in the system are analysed and incorporated in the switch specifications. The paper
will describe the measurements on device level and on the 10 kV assembly.The thermal management of the
whole switch and a mechanical layout will be presented. This work is being carried out under the EFDA
technology research programme activities, EFDA technology task TW3-THHE-CCGDS1, with financial
support from NWO
Corresponding Author:
A.B. STERK
sterk@rijnh.nl
FOM Institute for Plasma Physics p.o. box 1207 3430BE Nieuwegein THE NETHERLANDS
49
- B - Plasma Heating and Current Drive.
P3T-B-173
AN ALTERNATIVE ECRH FRONT STEERING LAUNCHER FOR THE
ITER UPPER PORT
RENE CHAVAN, MARK HENDERSON (1) FRANCISCO SANCHEZ (2)
(1)(2) Centre de Recherche en Physique des Plasmas, Association EURATOM - Confédération Suisse, Ecole Polytechnique
Fédérale de Lausanne, CH-1015 Lausanne, Switzerland
The purpose of the ITER electron cyclotron resonance heating (ECRH) upper port launcher will be to drive
current locally inside a q=3/2 or 2 island in order to stabilize the neoclassical tearing mode (NTM).
Unfortunately, the uncertainties due to our limited experience using ECCD for NTM stabilization magnified
by extrapolation to ITER, result in a relatively large range of current drive densities and injection angles that
may be needed on ITER. Although the remote steering (RS) launcher design offers the advantage of not
requiring moving parts within the vessel vacuum boundary (far from the thermal and nuclear radiation of the
plasma), it has a limited angular range and a relatively broad deposition at the resonance surface. A front
steering (FS) launcher offers an extended angular range and an increased current drive density relative to the
RS launcher. A FS launcher is already being planned for the equatorial port where thermal and radiation fluxes
are, in fact, higher than at the upper port. In light of this, an alternative FS launcher for application on the
ITER upper port is proposed, offering a wider steering angle (?±12?) and a higher ECRH power density than
the planned RS launcher. Neutron streaming calculations indicate that miter bends within the plug structure are
not required, and so launching systems can use straight waveguides with cross sections of ~16 cm2, which
simplifies the optical layout and reduces the space requirements for the internal components. The launcher is
capable of injecting over 8MW per port using a two mirror system (1 focusing and 1 steering) for focusing and
redirecting the beam towards the q=3/2 or 2 flux surfaces. The steering mechanism is bearing-free with flexure
pivots, in a compact cartridge capable of ±10? rotation (corresponding to ±20? for the microwave beam), with
cooling tubes coiled around the body for reducing stresses to levels corresponding to ITER design
requirements. A pneumatic seal-less actuator using helium integrated into the rotating mirror assembly offers a
fast and precise steering response (rotation control) avoiding push-pull rods and remote actuators. The result is
a complete self-contained frictionless kinematic assembly. The proposed design takes into account the specific
ITER requirements on operational reliability, remote assembly and handling. The complete design concept will
be presented along with a detailed comparison with the RS design.
Corresponding Author:
RENE CHAVAN
rene.chavan@epfl.ch
Centre de Recherche en Physique des Plasmas, Association EURATOM - Confédération Suisse, Ecole
Polytechnique Fédérale de Lausanne, CH-1015 Lausanne, Switzerland
50
- B - Plasma Heating and Current Drive.
P3T-B-174
DESIGN OF THE MM-WAVE SYSTEM OF THE ECRH UPPER
LAUNCHER FOR ITER
VERHOEVEN A.G.A. (TOON), W.A. BONGERS, A. BRUSCHI**, S. CIRANT**, I. DANILOV*, B.S.Q.
ELZENDOORN, J.W. GENUIT, M.F. GRASWINCKEL, R. HEIDINGER*, KASPAREK***, K. KLEEFELDT*, O.G.
KRUIJT, S. NOWAK**, B. PIOSCZYK*, B. PLAUM***, T.C. PLOMP, D.M.S. RONDEN AND H. ZOHM****
FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Nieuwegein, The Netherlands, *FZK,
Karlsruhe, **CNR, Milan, ***Univ Stuttgart, ****Max-Planck, Garching
The coordination of the design of the mm-wave system to be installed in the ITER Upper Ports is being carried
out at the FOM institute. The aim of the system is to inject Electron Cyclotron Waves (ECW) in the ITER
plasma in order to stabilize neoclassical tearing modes (NTM). Each upper-port launcher consists of eight mmwave lines capable of transmitting high power up to 2 MW at 170 GHz. In order to exploit the capability of
ECW for localized heating and current drive over a range of plasma radii in ITER, the ECH&CD upper port
launcher must have a beam steering capability. The steerable optic is considered a critical component for the
ECH&CD system and to avoid movable mirrors at the plasma-facing end of the launcher, the concept of
remote mm-wave beam steering (RS) is used, having a corrugated square waveguide within the launcher and
the steerable optic is then placed outside of the first confinement boundary of the vacuum vessel. Starting from
the gyrotrons mm-wave power will be transmitted towards the tokamak by circular evacu-ated waveguides.
Steering of the beam over a range of +/- 12 will be achieved by a mirror system consisting of a combination of
curved and rotating mirrors. Via the mirror system the beam will be directed into a square corrugated
waveguide. A single diamond-disk window and an isolation valve will provide the tritium boundary between
the pri-mary and secondary vacuum. At the end of the square waveguide, mm-wave beams will be guided
through penetrations in the front-shield blanket module by a fixed mirror towards the ITER plasma. This
mirror will have focusing properties in both directions. The resulting, effective steering range in the plasma is
still under study but will be around +/- 8 . The design analysis has demonstrated the feasibility of the remotesteering approach in the ITER envi-ronment. Now, the detailed design of the mm-wave layout has started
incorporating the remote-steering con-cept for the upper-port launcher and the aim is to come to a consistent
integration into the ITER environment. Furthermore, a full-scale mock-up line is being designed and built at
the appropriate ITER frequency, 170 GHz. Testing at the appropriate power level will start early 2005 at the
1.5 to 2 MW coaxial, short pulse gyro-tron at FZK, Karlsruhe. This work is being carried out under the EFDA
technology research programme activities, EFDA technology task TW3-TPHE-ECHULA and B1, with
financial support from NWO
Corresponding Author:
VERHOEVEN A.G.A. (TOON)
verhoeven@rijnh.nl
P.O. Box 1207, 3430 BE Nieuwegein, the Netherlands
51
- B - Plasma Heating and Current Drive.
P3T-B-175
DEVELOPING THE NEXT LHCD SOURCE FOR TORE SUPRA
KAZARIAN FABIENNE, B. BEAUMONT (1) E. BERTRAND (1) L. DELPECH (1) S. DUTHEIL (1) C. GOLETTO
(1) M. PROU. (1) A. BEUNAS (2) F. PEAUGER (2) PH. THOUVENIN (2)
(1) Association EURATOM-CEA, CEA/DSM/DRFC, CE Cadarache, 13108 St Paul lez Durance, France (2) THALES
ELECTRON DEVICES, 2 rue Latécoère. BP 23. 78141 Velizy cedex France
One of the main Tore Supra objectives is to produce long and performing discharges which studies are crucial
for the next step. A few years ago, the CIEL project has raised the power exhaust capability of the machine
and last year, a 6 minutes fully non inductive plasma has been sustained by 3 MW of LH power. The
performances of the pulses are now limited by the power injection level, and the CIMES project is targeting to
improve this point [1]. Associated to the manufacture of an ITER relevant PAM launcher [2], a new klystron is
under development at THALES ELECTRON DEVICES [3]. The upgrade will lead to an installed power of 12
MW in the lower hybrid transmitter. Each of the 16 tubes will work at 3.7 GHz, 76 kV, 22 A with an
efficiency up to 45 %. Several performances corresponding to different operating modes must be achieved: 700 kW CW mode on plasma (Value of Stationary Wave Ratio<=1.4), - 750 kW CW mode on matched load, pulsed mode on vacuum during antennas conditioning phase, - diode operation at full beam parameters. Each
mode presents its specific technological difficulties. On the other hand, the new klystrons will take place in our
existing installation which requires high compatibility with today equipment and induces fixed parameters. A
first breadboard has been manufactured and tested on the THALES test bed. It is now installed on Tore Supra
Lower Hybrid test bed. A second one is under finalization. Up to now, 757 kW peak at 76 kW, 22A (50 %
duty cycle) on breadboard 1 and 633 kW CW at 72.2 kV, 22A on breadboard 2 have been achieved on
matched load. A prototype, derived from this 2 models, will reach full performances in June. The tube’s
parameters are described in this paper. They are compared to those of the TH2103 presently settled in the
transmitter, their choices are detailed and explained. The developing phases, the results obtained with the
breadboards and the prototype as well as the technological difficulties are presented and analysed. [1] B.
Beaumont et al. Tore Supra Steady State Power and Particule Injection : the CIMES Project Fusion
Engineering and Design (56-57) (2001) [2] Ph. Bibet et al. ITER-Like PAM Launcher For TORE SUPRA
LHCD This conference [3] Ph. Thouvenin et al., High Power CW Klystron for Fusion Experiments IVEC
Conference (2004)
Corresponding Author:
KAZARIAN FABIENNE
kaza@drfc.cad.cea.fr
Association EURATOM-CEA, CEA/DSM/DRFC, CE Cadarache, 13108 St Paul Lez Durance Cedex
52
- B - Plasma Heating and Current Drive.
P3T-B-180
TOWARD AN LHCD SYSTEM FOR ITER
BIBET PHILIPPE, B. BEAUMONT, J. H. BELO, L. DELPECH, A. EKEDAHL, G. GRANUCCI, F. KAZARIAN, X.
LITAUDON, J. MAILLOUX, F. MIRIZZI, V. PERICOLI, M. PROU, K. RANTAMÄKI, A. TUCCILLO
On ITER, the LH system aims at supplying 20 MW CW for controlling the q-profiles that govern MHD
stability and confinement in partially (the so-called ‘hybrid’ regime at 12MA) and fully non-inductive steadystate operation (9MA). The designed LH system relies on a transmitter made of 24 (respectively 48) 5 GHz
1MW (500 kW) klystrons. These tubes are linked to one antenna based on the PAM (Passive Active
Multijunction) concept, via a 60 meters long oversized circular transmission line. The antenna geometry has
been chosen to radiate the wave with a spectrum having a N// index main value of 2 at a power density of 33
MW/m2. A common European effort including several associations (CEA, ENEA, UKAEA, IST, TEKES,
IPP-CZ) has been made in order to solve outstanding problems. The coupling in ITER scenario and
environment matters. Experiments performed on JET have shown the possibility to couple the LH wave in
environment similar to ITER. In order to verify the PAM concept, an antenna has been tested with success in
close collaboration between CEA and ENEA on the plasma of FTU at the end of 2003. The reliability of LH
for long pulse operation has been ascertained on Tore Supra where 370 s, 500 kA, one GJ fully non-inductive
discharges have been successfully obtained thanks to 3 MW of LH power. The transmission line components
and the antenna for ITER have been extensively studied within EFDA tasks. As a next step in demonstrating
ITER steady state scenarios and in order to routinely realise long pulse operation (1000 s), the Tore Supra LH
system is being refurbished within the CIMES project framework. The transmitter will be equipped with 16
klystrons 2103C from Thales. Their output power will be 700 kW for pulse length of 1000 s on VSWR smaller
than 1.4 at a frequency of 3.7GHz. Their development is a good milestone towards the design and the
realisation of a 500 kW CW 5 GHz tube. A new LH launcher based upon the PAM concept has been studied
and designed. It is made of 6 rows of eight 270 degrees bi junctions fed by TE•10 to TE•30 mode converters
that rely on the same concept than the one chosen for ITER launcher. The antenna is efficiently water-cooled,
in order to allow injecting 2.7 MW CW for a power density of 25 MW/m2. The chosen technology is similar
to the one envisaged for ITER. The new Tore Supra LH system will be available at the end of 2006. Its
achievement is an important step to confirm the implementation of a LH system on ITER.
Corresponding Author:
BIBET PHILIPPE
philippe.bibet@cea.fr
Association Euratom-CEA, CE Cadarache,13108, St Paul lez Durance, France
53
- B - Plasma Heating and Current Drive.
P3T-B-181
A N-PORT ERROR MODEL AND CALIBRATION PROCEDURE FOR
MEASURING THE SCATTERING MATRICES OF LOWER-HYBRID
MULTIJUNCTIONS
JOÃO P. S. BIZARRO,
In order to routinely test and measure the scattering (S) matrices of the multijunctions that build up lowerhybrid (LH) antennae, a crucial step to ensure that the launched spectrum is well defined and has a high
directivity for LH current drive, a method has been developed based on a generalization of the well-known
two-port error model and calibration procedure employed by comercially available network analysers. The
model presented takes into account the systematic errors inherent to microwave measurements (i.e. source and
load matching, reflection and transmission tracking, directivity, and isolation), which appear as error
coefficients determined via calibration standards. Furthermore, it allows for devices with any number of ports
and for the use of adaptors between the measuring system and the device under test, most probably needed in
the case of LH multijunctions, whose waveguides cannot be connected directly to coaxial cables.Once the
calibration has been completed, the whole S-matrix of a multijunction, considered as a multiple-port
microwave device, can be measured without having to carry out a long and monotonous series of operations,
making thus possible to reliably test the S-matrix of any arbitrary multijunction, at every stage of the
manufacturing process and as often as necessary.
Corresponding Author:
JOÃO P. S. BIZARRO
bizarro@cfn.ist.utl.pt
Centro de Fusão Nuclear, Associação Euratom-IST, Instituto Superior Técnico, 1049-001 Lisboa, Portugal
54
- B - Plasma Heating and Current Drive.
P3T-B-182
DESIGN AND FABRICATION OF THE "ITER-LIKE" SINGAP D¯
ACCELERATION SYSTEM
P MASSMANN, HPL DE ESCH, R S HEMSWORTH AND L SVENSSON
The SINGAP (SINGle APerture - SINgle GAP) acceleration concept is the simplified European alternative to
the Japanese Multi-Aperture, Multi-Grid (MAMuG) accelerator of the ITER Neutral Beam Injector (NBI)
reference design. To demonstrate ITER NBI (1 MV, 40 A) relevant beam optics in the Cadarache 1 MV, 100
mA test bed a maximum of the ITER SINGAP key parameters have been retained in the design of a new
“ITER-like” prototype accelerator, i.e. optimum D­­¯ current density at 1 MeV of 200 A/m², extraction, preacceleration and post-acceleration gaps as per the design for ITER. Because of their complexity the extraction
and pre-acceleration grid have been manufactured by electrolytic deposition of copper. The system is designed
to demonstrate also SINGAP "on to off-axis" beam steering by the simple transverse displacement of the postacceleration (SINGAP) electrode. Obtaining the current density level of >=200 A/m² is considered a crucial
part of the R&D. To maximize the probability of reaching this level two negative ion sources have been
developed. The first is a substantially revised, properly water-cooled, version of the prototype “Drift” Ion
Source [1]. Like the extraction and pre-acceleration grids, the side walls of this source, which feature different
size concentric rectangles of permanent magnet grooves and water channels, are fabricated by copper
deposition. The second source, the so-called “Alternative Source”, is a completely new design trying to
combine performance with ease of manufacture and low cost. Like the Drift Source, the Alternative Source is
immersed in vacuum, so that there are no vacuum tight seals on the source body. The side walls are made of
explosion bonded copper – stainless sandwich sheet material. Cooling channels are deep-drilled in the copper
layer, and the sheet is bent into an “L”-shape perpendicular to the water channels with the copper layer inside
the “L”. Two such “L’s” are put together to form the rectangular source body. The system is presently passing
its acceptance tests. In the paper we will present the details of the design, the fabrication methods and the
predicted performance. First results, which should be available by the time of the conference, will also be
given. REFERENCES [1] A Simonin, G Delogu, C Desgranges, M Fumelli, RSI 70 (1999) 4542
Corresponding Author:
P MASSMANN
massmann@pegase.cad.cea.fr
Association EURATOM - CEA CADARACHE, DRFC / SCCP, 13108 ST PAUL LEZ DURANCE Cedex, France
55
- B - Plasma Heating and Current Drive.
P3T-B-185
OPERATIONAL EXPERIENCE WITH UPGRADED JET NEUTRAL
BEAM INJECTION SYSTEMS
CIRIC DRAGOSLAV, CLIVE CHALLIS STEPHEN COX LEE HACKETT DAVID HOMFRAY DAVID KEELING
ROBERT KING IAN JENKINS TIMOTHY JONES ELIZABETH SURREY ADRIAN WHITEHEAD DAVID YOUNG
UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK
In the period 2001¡V2003, the JET Neutral Beam Injection (NBI) System has been upgraded with the design
goal of delivering 25 MW of deuterium beam power into the JET plasma. This major project involved the
following modification of the JET NBI System: „h Modification, pre-conditioning and installation of nine
130kV/60A Positive Ion Neutral Injectors (PINIs). Eight PINIs were installed on Octant 8 Neutral Injector
Box (NIB) and one on Octant 4 NIB in 2001 and 2002. „h Design, manufacturing and installation of new Box
Scrapers (in both NIBs) capable of handling higher power load. „h Re-configuration and commissioning of the
existing High Voltage Power Supplies (HVPS) to enable 130kV/60A operation for five upgraded PINIs. „h
Procurement, installation and commissioning of two new 130kV/130A HVPS units and corresponding control
systems to enable operation of four upgraded PINIs. All upgraded PINIs were conditioned without major
problems, with HVPS alarms being the most frequent fault condition. Five upgraded PINIs were
commissioned in 2002 and four PINIs, powered by the new HVPS modules, were brought into JET operation
in summer and autumn 2003. From November 2003 JET NBI System was operated using 16 PINIs for the first
time since the beginning of 2001. This lead to a record of 22.7MW of deuterium beam power injected into JET
plasma in January 2004. Although a new record in NBI heating power was established, the design value of
25MW could not be accomplished due to following reasons: „h Measurements of the neutral beam power
revealed that only 1.4MW (instead of 1.7MW) was delivered by one upgraded PINI. This power deficit could
be attributed to the reduction in the neutralisation target caused by the neutraliser gas overheating. „h The
operating voltage had to be limited to ~120kV (instead of 130kV) to prevent possible damage of the ion source
back-plates caused by back-streaming electrons ¡V one such event occurred in September 2002. Technical
improvements that are being carried out in the present JET shutdown (modification of the first stage neutraliser
and increase of the deceleration voltage) should enable the JET NBI System to operate at the 25 MW power
level in 2005. These improvements will be discussed in the paper, as well as some other issues related to full
power operation of the JET NBI System (pulse duration limits, beamline and torus protection, etc.). The work
is funded by EURATOM through the EFDA JET Operating Contract.
Corresponding Author:
CIRIC DRAGOSLAV
dragoslav.ciric@ukaea.org.uk
UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK
56
- B - Plasma Heating and Current Drive.
P3T-B-187
MAST NEUTRAL BEAM LONG PULSE UPGRADE
GEE STEPHEN, ANDREW BORTHWICK DRAGOSLAV CIRIC GEORGE CRAWFORD LEE HACKETT DAVID
HOMFRAY DAVID MARTIN JOSEPH MILNES TIM MUTTERS MARTIN SIMMONDS RICHARD SMITH PAUL
STEVENSON CHRIS WALDON SIMON WARDER ADRIAN WHITEHEAD DAVID YOUNG
UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK
Neutral beam heating is the main auxiliary plasma heating system on the Mega Amp Spherical Tokamak
(MAST) at Culham. Until summer 2003, experiments on MAST were carried out using relatively short (200400 ms) plasma pulses. Two Neutral Beam Injectors (NBI), each equipped with one duopigatron ion source
(on loan from Oak Ridge National Laboratory), were delivering up to 3 MW of deuterium neutral beam power
into the MAST plasma for the duration of up to 300 ms. During the recent shutdown, new components (central
solenoid, divertor, etc.) were installed to enable long pulse operation (up to 5s) of the MAST machine. To
accommodate the long pulse operation requirement, the NBI system is also being upgraded to deliver up to 5
MW of deuterium neutral beam power into the MAST plasma, for the duration of up to 5 seconds. Two
duopigatron ion sources are being replaced with the JET type Positive Ion Neutral Injectors (PINIs). The
MAST PINI design is a modification of the JET high current tetrode injector, with nominal deuterium beam
voltage and current of 75kV and 65A, respectively. Each injector will deliver up to 2.5 MW of deuterium
neutral beam power for up to 5 seconds. In addition to the replacement of the two injectors, the majority of the
components of the MAST NBI system are being replaced or modified. Each beamline is now equipped with
new, hypervapotron based calorimeters and residual ion dumps capable of handling long pulse/high power
beams. They are instrumented with ~100 thermocouples (per beamline) to enable beam characterisation. Most
of the high voltage power supplies and controls are being modified or replaced to allow long pulse operation.
Some of the new features are high voltage regulation, re-application and modulation. The new beam interlock
system is being installed to protect both beamline and MAST vessel components from the excessive beam
power loading during fault conditions (over-pressure, magnet current mismatch, low plasma density, etc.).
Data acquisition system and timer controls are also being upgraded to allow fast collection and storage of
increased number of signals for considerably longer duration. The first MAST PINI will be brought into
operation in summer 2004 and the second one at the beginning of 2005. The paper will address various design
issues and the initial operational experience with upgraded MAST NBI system. Work partly funded by
EURATOM and the UK Engineering and Physical Sciences Research Council
Corresponding Author:
GEE STEPHEN
stephen.gee@ukaea.org.uk
UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK
57
- B - Plasma Heating and Current Drive.
P3T-B-188
THE ITER NEUTRAL BEAM TEST FACILITY : DESIGNS OF THE
GENERAL INFRASTRUCTURE, CRYOSYSTEM AND COOLING
PLANT
CORDIER JEAN-JACQUES, R. HEMSWORTH (1), M. CHANTANT (1), B. GRAVIL (1), D. HENRY (1), F.
SABATHIER (1), L. DOCEUL (1), E. THOMAS (1), D. VAN HOUTTE (1), P. ZACCARIA (2), V. ANTONI (2), S.
DAL BELLO (2), A. MASIELLO (2), D. MARCUZZI (2), M. DREMEL (3), C. DAY (3)
(1) CEA DSM / Département Recherche Fusion Contrôlée, CEA/Cadarache, 13108 Saint Paul Lez Durance Cedex, France
(2) CONSORZIO RFX, Corso Stati Uniti 4, 35127 Padova Italy (3) FZK, Institut für Technische Physik, Karlsruhe 76021,
Germany
In the frame an EFDA contract (task ref. TW3-THHN-IITF1) the CEA, in close collaboration with the
Consorzio RFX, Padua, and FZK, Karlsruhe, is carrying out a design of the ITER Neutral Beam Test Facility
(NBTF). The main objective is to demonstrate its reliability and to optimise the performances of the main
beam line components during operation, i.e. the beam source, the neutraliser, the residual ion dump, and the
calorimeter. The proposed design of the Neutral Beam Test Facility general infrastructure layout is described
in the paper, with taking into account the associated safety requirements (Neutrons and X-ray production). The
infrastructure includes integration studies of the cooling plant, the cryosystem and the forepumping system.
The ITER neutral beam heating and current drive system is equipped with a cryosorption (activated charcoal)
cryopump made up of 12 panels, refrigerated in parallel by 4.5 K, 0.4 MPa supercritical helium. The pump is
submitted to a non homogeneous flux of H2 or D2 gas and the absorbed flows vary from 3 Pa.m-3.s-1 to 35
Pa.m-3.s-1. The NBTF also requires a cryosystem to supply the necessary cryogens to the cryopump. The 4.5
K cryopanels must be periodically regenerated at 90 K and, occasionally, at 470 K. The cool-down times from
room temperature and after regeneration depend strongly on the refrigeration capacity. Regeneration and cooldown phases of the cryopanels are evaluated for the test facility operation. The consequences of an optimised
4.5 K cold power and 80 K helium gas refrigerators on the operation plan have been analysed and will be
discussed. A total power of about 50 MW will have to be removed in steady state during the two stages short
and long pulse operation of the NBTF. The cooling plant and the associated pressurised water loops that are
required for cooling down the high voltage components (beam source, accelerator grid, transmission line, and
HV bushing) and the low voltage components (neutraliser, residual ion dump, calorimeter) are designed for
both the short (20 s), and long operating pulses (3600 s) that are to be demonstrated on the test facility. The
paper describes the design and the characteristics of both the optimised Primary Heat Transfer System (PHTS)
and the associated Heat Removal System (HRS). A comparison is made between the cryosystem and water
cooling systems proposed for the NBTF and the corresponding ITER NBI heating system reference design.
Corresponding Author:
CORDIER JEAN-JACQUES
jean-jacques.cordier@cea.fr
Association EURATOM-CEA, DSM / Département Recherche Fusion Contrôlée, 13108 Saint Paul Lez Durance
Cedex France
58
- B - Plasma Heating and Current Drive.
P3T-B-201
PROGRESS OF THE KSTAR ICRF COMPONENTS DEVELOPMENT
FOR LONG PULSE OPERATION
B.G. HONG, Y.D. BAE, C.K. HWANG, J.G. KWAK, S.J. WANG AND J.S. YOON
The ICRF system for the KSTAR tokamak [1] is being developed to support long pulse, high beta, advanced
tokamak physics experiments. The system will provide a function of pressure and current density profile
control by providing heating and on-axis/off-axis current drive over a range of magnetic fields with the
frequency range of 25-60 MHz. And it will deliver 6 MW of RF power to plasma from 2009 with long pulse
lengths operation capability up to 300 second. To transmit MW level of RF power for a long pulse, ICRF
components such as antenna, vacuum feedthrough, and tuning components should have the high stand-off
voltage and current without breakdown, and operational reliability. A high power density (~ 1 kW/cm2) ICRF
antenna and a vacuum feedthrough which has two alumina (Al2O3, 97%) ceramic cylinders and O-ring seal
have been developed. High power RF tests were performed with the antenna installed in the RF test stand. The
peak voltages over 35 kVp for 300 second were found. Tuning components which use silicon oil (relative
dielectric constant, 2.74) as insulating medium were developed for long pulse operation. They have a high
stand-off voltage (> 40 kV) and can be used for matching during a shot by changing the level of silicon oil.
Feasibility study for a coaxial fast ferrite tuner where the space between the conductors is partially filled with
coaxial ferro-magnetic materials is also under investigation for matching large and fast changes.of the load,
and the results are reported. The results of the development will be applicable for the long pulse, high power
operation of the KSTAR ICRF system. [1] G.S. Lee et. al., “The KSTAR Project: Advanced Steady-State
Superconducting Tokamak Experiment”, Nuclear Fusion 40 (2000) 575-582.
Corresponding Author:
B.G. HONG
bghong@kaeri.re.kr
P.O. Box 105, Yusong, Daejeon, 350-600, Korea
59
- B - Plasma Heating and Current Drive.
P3T-B-204
RECENT PROGRESS OF NEGATIVE ION BASED NEUTRAL BEAM
INJECTOR FOR JT-60U
UMEDA NAOTAKA, YAMAMOTO TAKUMI LARRY GRISHAM(1) KAWAI MIKITO OHGA TOKUMICHI AKINO
NOBORU MOGAKI KAZUHIKO YAMAZAKI HARUYUKI KIKUCHI KASTUMI JT-60 NBI TEAM
Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukouyama Naka-machi Naka-gun
Ibarakiken, 311-0193 Japan (1)Princeton Plasma Physics Laboratory, Po Box 451, Princeton, N.J. USA 08543
The 500keV negative ion based neutral beam injection (N-NBI) system for JT-60U was constructed in 1996,
and thereafter has been in operation for study of core plasma heating and non-inductive current drive. Some
modifications of negative ion source have been recently conducted so as to expand pulse duration to 30 sec
from 10 sec, which is design value. Heat load on the grounded grid in the ion source was higher than the
design value by three times and the temperature of water cooling for the grounded grid increased up to 90
degree. It was difficult to inject beam for the long time which the acceleration grids reached to thermal steady
state. On the other hand beam limiters at NBI port are not cooled forcedly and then those temperatures increase
with beam pulse duration. It is also important to reduce the heat load on the limiters in order to expand beam
pulse. From the calculation of the beam trajectory, beams extracted from edge grid segments deposit largely on
the limiters than inside segments. In order to lessen the heat load on the grounded grid and on the beamline
limiters, outermost segments of the plasma grid extracting negative ions were masked and all the acceleration
grid segments of the down stream of the masked segments were altered to the grids which had large hole to
exhaust gas. By these modifications gas conductances of extractor have decreased and those of accelerator
have increased. The gas pressure of the arc chamber was kept around 0.3Pa and the pressure of the extractor
and the accelerator diminished about 30%. As a result striping loss of negative ion was simulated to diminish
by 20%. From the measurement of the heat load of acceleration grids and beam line components, the ratio of
grounded grid heat load to beam power diminished from 0.08 to 0.06 in the optimum condition and the
maximum acceleration efficiency increase 0.71 to 0.79. Temperature rise of beam limiter has decreased by
35%. As a result the long pulse injection for 17 sec with 1.6MW power has been achieved at 366keV beam
energy.
Corresponding Author:
UMEDA NAOTAKA
umedana@fusion.naka.jaeri.go.jp
Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1, Mukouyama, Nakamachi, Naka-gun, Ibarakiken, 311-0193, Japan
60
- B - Plasma Heating and Current Drive.
P3T-B-210
TESTS AND FIRST RESULTS OF A LOAD RESILIENT ICRH ANTENNA
ON TEXTOR
VERVIER MICHEL, P. DUMORTIER S. GRINE A. MESSIAEN G. VAN WASSENHOVE
Trilateral Euregio Cluster Laboratoire de Physique des Plasmas / Laboratorium voor Plasmafysica, EURATOM Association,
B-1000 Brussels, Belgium
Due to rearrangement of the diagnostic positions resulting from DED installation on TEXTOR, a new antenna
system has been installed to be compatible with the inlet of the diagnostic beam between its two radiating
straps. As described in [1] this antenna has been designed to test the “conjugate-T” mode of operation which is
foreseen to solve the problem of generator tripping occurring during the ICRF heating of Elmy H-mode
plasmas. But this antenna is also able to operate in the conventional way with pi or 0 phasing. The paper
describes the installation of the antenna system. It consists of a toroidal pair of resonant straps, each strap
being ended by a variable vacuum capacitor and fed by means of a tap. The "conjugate-T" mode of operation
is obtained by an appropriate de-tuning of each resonating circuit strap-condenser and by means of the
adjustment of the feeding line length between the tap and the “T”. The paper deals with the calibration of the
antenna and line system at low power in order to allow detailed measurement of the coupling characteristics
and to ensure the protection of the condensers against over-current. It describes also the analysis of the tuning
procedure of the conjugate T and of the deduced practical method to optimize its performances. The mutual
coupling between the two straps can reduce the performances of the conjugate-T. This problem is also
analyzed. Diagnostics by means of current and voltage probes and directional couplers have been installed on
the antenna system and on its feeding line. The change of phase difference between the straps which enables
the load resilience is also measured. The paper will present the first results on plasma using these data and
from the modeling of the complete antenna system. [1] F. Durodié et al., “Development of a load-insensitive
ICRH antenna system on TEXTOR”, proceedings of 22nd SOFT, Helsinki 2002, pp. 509.
Corresponding Author:
VERVIER MICHEL
Michel.Vervier@rma.ac.be
30 av. de la Renaissance, B-1000 Bruxelles, Belgium
61
- B - Plasma Heating and Current Drive.
P3T-B-211
REALISATION OF A TEST FACILITY FOR THE ICRH ITER PLUG-IN
BY MEANS OF A MOCK-UP WITH SALTED WATER LOAD
MESSIAEN ANDRÉ, P.DUMORTIER R.KOCH P.LAMALLE F.LOUCHE J.L.MARTINI M.VERVIER
Trilateral Euregio Cluster Laboratoire de Physique des Plasmas / Laboratorium voor Plasmafysica, EURATOM Association,
B-1000 Brussels, Belgium
A conceptual design of a 20MW ICRH plug-in for ITER in the frequency band 40-55MHz with external
matching has been developed [1]. The main advantages of this design are the absence of in-vessel remotely
operated components to achieve the matching and the use of passive junctions which minimises the number of
matching circuits. Indeed the 24 straps of the radiating array are grouped in 4 conjugate T circuits in order to
provide the highly load resilient matching needed in presence of ELMy discharges. The straps will
unavoidably be coupled to each other as they are radiating in the same medium. The load resilience and the
theoretical expectations of the effects of such coupling have to be checked before the installation of the
antenna array on ITER with a good simulation of the plasma load. Tests in absence of plasma are useful but
will not at all simulate the electromagnetic properties in presence of plasma nor allow testing the tuning
algorithm. The first part of the paper shows that: (i) test with realistic plasma-like load conditions can be
obtained with a large dielectric constant medium facing the strap array, (ii) when decreasing the length and
increasing the frequency by the same scale factor the impedance matrix of the array remains identical, (iii)
salted water can advantageously be used as a load. The second part of the paper describes the construction of
the mock-up of the complete antenna array (with a scale-down factor of 5), of its feeding by passive 4-port
junctions and of its water load. The addition of salt in the water avoids the use of a large tank and allows
adjusting the loading properties. Measurements in the frequency range 200-275MHz provides identical
impedance matrix as the full scale system and can be directly compared with modelling obtained from the CST
Microwave Studio (MWS) software. Resilience of the full-scale system to ELMs can be checked on the mockup by varying the distance array-water load. The mock-up also allows testing tuning algorithms in presence of
mutual coupling and the use of polychromatic heating to decrease the effects of this coupling. [1] P.Dumortier
et al., Final report on Task FU05-CT 2002-00094 (EFDA/02-675), LPP-ERM/KMS Int. Rep. 121, A.Messiaen
et al. “Radio-frequency power in plasmas” ( Proc. 15th Top. Conf. On Radio-Frequency Power in Plasmas,
Moran, Wyoming, May2003) AIP conf proceedings volume 694 p.142.
Corresponding Author:
MESSIAEN ANDRÉ
Andre.Messiaen@rma.ac.be
30, av. de la Renaissance, B-1000 Brussels, Belgium
62
- B - Plasma Heating and Current Drive.
P3T-B-218
STUDY OF MUTUAL COUPLING EFFECTS IN THE ANTENNA ARRAY
OF THE ICRH PLUG-IN FOR ITER
P. U. LAMALLE, A.MESSIAEN P.DUMORTIER F.LOUCHE
Trilateral Euregio Cluster Laboratoire de Physique des Plasmas / Laboratorium voor Plasmafysica, EURATOM Association,
B-1000 Brussels, Belgium
The ICRH launcher proposed for ITER is constituted by a large number (presently 24 for [1] and [2]) of short,
closely packed radiating straps in order to decrease the antenna voltage. To insure the compatibility with the
Elmy H-mode operation of ITER the antenna system must be insensitive to large load variations. This is
achieved by grouping the straps in several “conjugate-T” (CT) matching systems. The tuning is performed by
means of vacuum capacitors [1] or line stretchers [2]. The conceptual design has been made without
considering the mutual coupling between the radiating straps. However, since they are radiating in the same
medium, the latter are unavoidably coupled. This coupling influences the load resilience performance,
considerably increases the complexity of the simultaneous tuning of various CT’s, creates significant voltage
imbalances between straps, and can even result in power transfer between the different power sources feeding
the array. In order to provide realistic loading conditions for the study of matching, the first part of the paper
describes the properties of the impedance matrix of the array. In particular, it is shown that the mutual
impedances have an important resistive component. The second part describes the effect of the mutual
impedance on one CT circuit. It shows that a large load resilience can still be obtained, but that the matching
conditions are more critical and that the reactive part of the mutual coupling can lead to large unbalance and
phase variation between the radiated power by the two parts of the CT. Remedies and a first practical tuning
method are proposed. The third part deals with the problem of the coupling between the different CT’s and
their power sources. It underlines the high complexity of the simultaneous tuning because the degeneracy of
the tuning values of the capacitors or line stretchers of the different CT’s is lifted by the mutual coupling.
Practical tuning algorithms and the possible use of ‘polychromatic’ heating (i.e. the operation of different parts
of the array at slightly different frequencies) to alleviate the adverse effects of mutual coupling are discussed.
[1] Detailed Design Description Ion Cyclotron Heating and Current Drive System WBS 5.1 (DDD). [2]
P.Dumortier et al., Final report on Task FU05-CT 2002-00094 (EFDA/02-675), LPP-ERM/KMS Int. Rep. 121
Corresponding Author:
P. U. LAMALLE
Philippe.Lamalle@rma.ac.be
30, av. de la Renaissance, B-1000, Brussels, Belgium
63
- B - Plasma Heating and Current Drive.
P3T-B-221
STATUS AND PLANS FOR THE DEVELOPMENT OF AN RF
NEGATIVE ION SOURCE FOR ITER NBI
FRANZEN, PETER, H. D. FALTER, M. BANDYOPADHYAY, U. FANTZ, B. HEINEMANN, W. KRAUS, P.
MCNEELY, R. RIEDL, E. SPETH, A. TANGA, R. WILHELM
Selected also for oral presentation
O3A-B-221
The reference design for the neutral beam injection system of ITER is based on arc sources rated for 40 A of
D- ions extracted from a 1.5 x 0.6 m2 source with a net extraction area of 0.2 m2. The main problem of the arc
source is the limited lifetime of the filaments. Furthermore it is suspected that the arc current is responsible for
the source non uniformity observed in large arc sources for negative ion production. Therefore RF sources,
developed successfully at IPP for neutral beam heating based on H+ and D+ ions, offer substantial advantages
for ITER neutral beam heating. The development of an RF ion source for negative ions has been carried on at
IPP since December 2002 within the framework of an EFDA contract. So far current densities of 260 A/m2 for
hydrogen and 170 A/m2 for deuterium have been achieved for an extraction area of 0.007 m2 at a source
pressure of <0.5 Pa. Caesium evaporation is necessary for these high negative ion yields. The electron/ion ratio
can be kept below 1 for both hydrogen and deuterium by biasing the plasma grid against the source body with
10-20 V if the filter field, i.e. a magnetic field above the plasma grid which suppresses the electrons, is
sufficiently strong. Deuterium requires a stronger filter field than hydrogen. However, the useful RF power is
limited by the strong filter field with the present set-up. Modifications to overcome this limitation are being
prepared. An extension of the extraction area from 0.007 m2 to 0.015 m2 has already been demonstrated
without loss of current density. Parallel to the source development the design and manufacturing of a test
facility for pulses of up to 1 hour duration is proceeding, scheduled for commissioning towards the end of
2004. A scaled up ion source with the same width and half the length of the ITER reference source will
become available for commissioning early in 2005. The paper will present as a summary an overview of the
latest results of the source development, of the design of the half size ITER source and of the status of the long
pulse development. The details will be presented in several other papers.
Corresponding Author:
FRANZEN, PETER
peter.franzen@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Postfach 1533, D-85740 Garching, Germany
64
- B - Plasma Heating and Current Drive.
P3T-B-225
DEVELOPMENT AND CONTRIBUTION OF RF HEATING AND
CURRENT DRIVE SYSTEMS TO LONG PULSE, HIGH PERFORMANCE
EXPERIMENTS IN JT-60U
SHINICHI MORIYAMA, MASAMI SEKI, SHUNSUKE IDE, AKIHIKO ISAYAMA, TAKAHIRO SUZUKI, TSUNEYUKI
FUJII AND JT-60 TEAM
Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukohyama, Naka-machi, Naka-gun,
Ibaraki-ken, 311-0193 Japan
Selected also for oral presentation
O3A-B-225
The recent experiment campaign of JT-60U was started in November 2003 with emphasis on long sustainment
of high performance plasmas. The maximum duration of the plasma, which was 15 sec, has been extended to
65 sec by means of modification of control systems and saving volt-sec consumption by RF and NB heating
and current drive. The major purposes of this experiment campaign are; 1) long sustainment of high boot-strap
current fraction, 2) long sustainment of high-beta plasma, 3) improvement of quasi steady state beta by
suppression of the neoclassical tearing mode (NTM). These are important issues to the reactor. The electron
cyclotron (EC) and lower hybrid (LH) heating and current drive systems play important roles in these
challenges. For improvement of confinement by current profile control or by NTM suppression, EC system is
effective. Movable antennas can steer beam angle to put current drive location at the mode island by real-time
feedback control. The target of the EC operation in long pulse is 0.6 MW for 30 sec with 4 gyrotrons, though
10 MJ (2.8MW, 3.6sec) was recorded in high power operation before 2003. One of the critical issues for the
long pulse operation is detuning due to decay in collector current of the gyrotron. The decay comes from the
heater cooling by continuous electron emission. As a countermeasure for this issue, active adjustments for the
heater current and anode voltage during or just before the pulse have successfully extended the duration of a
good oscillation condition for the gyrotron. A "waveguide dummy load" for steady state 1MW absorption is
used in these trials. Improvement in cooling of the transmission components and efforts in noise suppression
have enabled the long pulse operation. As a result, 0.4 MW for 16 sec with 1 gyrotron has been achieved in
March 2004. LH system is effective for current drive and is a key to extend pulse duration of reversed shear
plasmas in this experiment campaign. In the LH system, the klystron was adjusted for long pulse, and the
antenna mouth was newly implemented with carbon grill. The original metal antenna mouth had been partially
deformed by heat load from the plasma and the RF arcing for 10 years' operation. The power handling
capability and the durability for heat loads are expected to be improved by the carbon-grill-antenna.
Conditioning of the antenna is under going and injection of 0.9 MW (5.1 MJ) has been achieved by March
2004.
Corresponding Author:
SHINICHI MORIYAMA
moriyama@naka.jaeri.go.jp
RF Facilities Division, Department of Fusion Facilities, Naka Fusion Research Establishment, Japan Atomic
Energy Research Institute, 801-1 Mukohyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan
65
- B - Plasma Heating and Current Drive.
P3T-B-227
RF-SOURCE DEVELOPMENT FOR ITER: LARGE AREA H- BEAM
EXTRACTION, MODIFICATIONS FOR LONG PULSE OPERATION
AND DESIGN OF A HALF SIZE ITER SOURCE
KRAUS, WERNER, B. HEINEMANN, H. D. FALTER, U. FANTZ, T. FRANKE, P. FRANZEN, D. HOLTUM, CH.
MARTENS, P. MCNEELY, R. RIEDL, E. SPETH, R. WILHELM
At IPP RF ion sources are developed for the ITER neutral beam heating since 2002 through an EFDA contract.
While most of the physical experiments are carried out with a net extraction area of 74 cm2 on the “Batman”
testbed, on a second test facility (multi ampere negative ion test unit “Manitu”) the experiments are focussed
on large area extraction and long pulses. In a first step the extraction area has been extended to 152 cm2 (300
apertures, Ø 8 mm). For the HV power supply a novel HV circuit has been commissioned, which utilizes two
switching tubes for the generation of the extraction and the acceleration voltage. After a stable surface
production of negative ions has been achieved in the cesiated source, it delivers very reproducible high Hcurrent densities, being almost independent on the filling pressure. At 0.45 Pa with 85 kW RF power a
calorimetrically measured H- current density of 20 mA/cm2 has been reached, which is consistent with the
results obtained with the small extraction area. The addition of argon reduces the ion current considerably. In a
second step the extraction area has been enlarged to 300 cm2 which is about the area supplied by one RF
driver in the ITER size source. To demonstrate the current density and plasma homogeneity over the whole
ITER extraction area, a half size ITER source has been designed and is under construction. It will have the
total width and half of the length of the ITER source (800 x 900 mm2). Two 180 kW RF power supplies, and a
dummy plasma grid simulating the ITER gas conductivity are foreseen. Single hole extraction and Faraday cup
measurements are planned on about 20 apertures. Furthermore one testbed will be upgraded to demonstrate cw
operation in D- with several 3600 s pulses in spring 2005. This requires replacing the existing titanium
evaporation pumps by cryo pumps developed by FZK and installing a new calorimeter suitable for 360 kW
total power and a maximum power density of 0.6 kW/cm2. New power supplies for beam extraction
(15kV/35A), acceleration (35kV/15A) and RF (180 kW) are necessary as well as an upgrade of the data
acquisition and cooling system. This paper will describe the results of the beam extraction experiments, the
design of the half size ITER source and the modifications of the main components for long pulse operation.
Corresponding Author:
KRAUS, WERNER
kraus@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, 85748 Garching, Germany
66
- B - Plasma Heating and Current Drive.
P3T-B-229
DIAGNOSTICS AND MODELING OF THE PLASMA IN BATMAN
RADIO FREQUENCY ION SOURCE
TANGA ARTURO, M.BANDYOPADHYAY, H. FALTER, U. FANTZ, P. FRANZEN, B. HEINEMANN, W. KRAUS, P.
MCNEELY, R. RIEDL, E. SPETH AND R. WILHELM
Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmann str. 2, D-85748, Garching, Germany.
This paper describes the development of the diagnostics and computational activities for the negative hydrogen
ion source for the neutral beam system for ITER done at IPP, Garching. Radio frequency (RF) sources have
advantages of low maintenance and more operational time availability compared to the arc sources. Diagnostic
measurements on the other hand have to face the difficulties of RF pick up and the modulation of the electron
population. Plasma potential, density and temperature profiles are routinely obtained using a Langmuir probe,
while a Mach probe has been used to provide the Mach number as well as the pattern of the plasma flow in the
ion source. Measurements in the region of the plasma grid show the effect of bias as well as the pattern of the
fields which determine the initial orbits of the extracted particles. The measured fluid motion is amenable to
fluid dynamic analysis which has been done along the axis of symmetry. From the results of the analysis it is
shown that the addition of a transverse magnetic field reduces strongly the plasma flow velocity. Modulation
of plasma parameters have been used to produce accurate measurements using phase sensitive techniques. The
combination of such experimental data with a Monte-carlo code for the treatment of neutrals, molecules and
individual ions will help further to predict the performance in the development of a full size ITER source.
Corresponding Author:
TANGA ARTURO
att@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmann str. 2, D-85748, Garching,
Germany.
67
- B - Plasma Heating and Current Drive.
P3T-B-246
ECH MW-LEVEL CW TRANSMISSION LINE COMPONENTS
SUITABLE FOR ITER
OLSTAD, R.A., J.L. DOANE (1), C.M. MOELLER (1)
(1) General Atomics, P.O. Box 85608, San Diego, California 92186-5608
The ECH transmission lines for ITER will require performance parameters not yet entirely demonstrated in
ECH systems on present magnetic fusion energy machines. The key performance requirements for the main
ITER transmission lines are operation at 1 MW for pulse lengths of 400 s up to 3600 s (essentially cw) at a
frequency of 170 GHz. An additional consideration for transmission line performance is the possibility that
ITER will use 2 MW coaxial cavity gyrotrons currently under development by Forschungszentrum Karlsruhe
(FZK) and other European Associations and European tube industry. This paper addresses the progress made
by General Atomics in the various transmission line components suitable for use on ITER at 170 GHz, as well
as at 120 GHz for plasma startup. ITER design documents call for a corrugated waveguide inner diameter of
63.5 mm; many components have already been fabricated in this diameter, and those that have been made in
other diameters (namely 31.75 mm and 88.9 mm) can readily be modified to a 63.5 mm i.d. design. In some
cases, water cooling must be added to present designs to remove heat deposited during cw operation of the
components. In addition to the main transmission lines, there are corrugated waveguide components
incorporated into the ECH launcher systems (equatorial and upper launchers). The status of the development
of these components, including remotely steerable launcher components, is also presented. This paper focuses
on those components needing design modifications to meet ITER requirements (i.e. frequency, power level,
pulse length, diameter). The heat loads and resultant temperature increases for critical components are
estimated. For those components whose temperatures will exceed safe limits, design changes already
underway or planned will be addressed. Components being designed for ITER and other cw applications
include Matching Optics Units (MOUs), aluminum waveguide sections adjacent to miter bends, compact
dummy loads, dc breaks, waveguide bellows, stainless steel waveguides, and remote steering launcher
waveguides.
Corresponding Author:
OLSTAD, R.A.
olstad@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
68
- B - Plasma Heating and Current Drive.
P3T-B-267
STATUS OF THE TJ-II ELECTRON BERNSTEIN WAVES HEATING
PROJECT
FERNÁNDEZ ÁNGELA (1), KAREN SARKSYAN (2) ÁLVARO CAPPA (1) FRANCISCO CASTEJÓN (1) NICOLAI
MATVEEV (3) ÁNGELA GARCÍA (1) MERCEDES MEDRANO (1) JOHN DOANE (4) CHARLES MOELLER (4)
JOSÉ DONCEL (1) ANTONIO PARDO (1) MAXIM TERESHCHENKO (2) NICOLAI KHARCHEV (2) ALEXANDER
TOLKACHEV (1)
(1) EURATOM-CIEMAT Association. Madrid, Spain (2) General Physics Institute, Moscow, Russia (3) State Unitary
Enterprise, Moscow, Russia (4) General Atomics. San Diego, California, USA
The present status of the main components of the TJ-II Electron Bernstein Waves (EBW) heating system and
the theoretical calculations performed to determine the precise launching and beam structure conditions are
presented.The O-X-B scenario has been chosen for first harmonic (28 GHz) heating of an overdense plasma.
One 300kW-gyrotron (cathode voltage: 60-70kV, current: 13-25 A, pulse length: 100ms), which was used for
ECR heating in TJ-IU torsatron, has been checked and is ready for installation in its cryomagnet. The design of
a new high voltage power supply unit, which provides the formation of a stabilized negative voltage pulse up
to 70 kV and a maximum current of 25 A, is finished. The assembly and installation should be completed at
the beginning of 2005. The microwave power will be transmitted by an oversized corrugated waveguide
(length: 7 m, two continuous curvature bends with an estimated overall transmission loss of about 2 to 3%,
inner diameter: 45 mm, operation at atmospheric pressure). Two ellipsoidal mirrors are necessary to optimise
the Gaussian beam parameters at the input of the waveguide to achieve minimal matching losses. Two
corrugated mirrors are used to get the optimal polarization, so that the highest EBW absorption efficiency can
be achieved. A movable internal mirror is needed in order to focus the beam and to accomplish the restrictive
launching angle conditions. The support and its handling is being design and will be finished when the
theoretical calculations confirm the optimal position and beam shape. The present cooling system of the two
53.2 GHz-gyrotrons of the ECRH system is being upgraded to cool the 28 GHz-complex. The extension of the
system will include the water supply for the gyrotron, the HV power supply and the calorimetric system. On
the primary circuit, an additional pump will be installed to supply the different components in parallel circuits,
meanwhile the cooling power of the current plate heat exchanger will be increased suitably. To measure the
power, a calorimeter with teflon pipes will be installed in front of the gyrotron window. A power monitor will
be installed in the waveguide. This element is important to perform power modulation experiments to obtain
the EBW power deposition profile. The start of the experiments is schedule for 2005.
Corresponding Author:
FERNÁNDEZ ÁNGELA (1)
angela.curto@ciemat.es
Association EURATOM-CIEMAT. Avda. Complutense, 22. 28040 Madrid.Spain
69
- B - Plasma Heating and Current Drive.
P3T-B-283
THE ASDEX UPGRADE ICRF SYSTEM: OPERATIONAL EXPERIENCE
AND DEVELOPMENTS
FAUGEL HELMUT, P. ANGENE, W. BECKER, F. BRAUN, B. ECKERT, F. FISCHER, G. HEILMAIER, J.
KNEIDL, J.-M. NOTERDAEME, G. SIEGL, E. WUERSCHING
The ICRF system on ASDEX Upgrade (AUG) consists of four generators with 2 MW each from 30 to 80
MHz, declining to 1 MW at 120 MHz, four two stub matching systems and four two strap antennas. The length
of the antenna feeding lines allows matching at four frequencies: 30, 36.5 and 40.7 MHz, used for H minority
in the 2 to 2.5 T range, and 61.7 MHz for second harmonic near 2 T. The phasing of the antenna straps is set to
0, pi. At 30 MHz, the system can be switched to asymmetric phasing for two antennas in the co-current and
two antennas in the counter-current direction. ICRF has been operational on AUG since 1992. First tests in
1996 using 3 dB hybrids on two generators led to there installation on all four generators in 1998. This made
operation with type I ELMs possible. ICRF has since become a reliable and powerful heating system on AUG
under all conditions. The increased reliability of the ICRF further comes from: - intensive conditioning after
each vent - a new system to switch off the generators - repeated use on plasma. The standard use of 4 x 0.8
MW on the first plasma shot of each day provides a renewed on-plasma conditioning - a matching program to
calculate the matching ICRF delivered pulses with up to 7 s length, a maximum RF power of 7.2 MW (90% of
the installed generator power) and an energy of 38 MJ. Present developments aim at using the ICRF heating
increasingly in feed-back control of the discharge parameters, e.g. to keep the plasma energy constant. The
huge variation of the generator output power does however raise technical problems, such as a high power
dissipation of the final stage tube or a too high screen grid current. This problem can be avoided by controlling
the anode voltage. In a first test on a dummy load, the anode voltage of the final stage was set to 14 kV for
zero output power, increasing to 23 kV at 2 MW. The overall performance of these tests were much better than
with a fixed anode voltage of 23 kV. In this case the anode power dissipation exceeded the 1250 kW limit at
about 700 kW RF resulting in a switch off of the generator. Experiments will show if the anode voltage control
can be implemented on plasma discharges. Longer term development to use ICRF beyond heating would
benefit from an increased flexibility in the choice of frequency and phasing and from an improved antenna
spectrum (using 4 straps).
Corresponding Author:
FAUGEL HELMUT
faugel@ipp.mpg.de
Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr.2, D-85748 Garching, Germany
70
- B - Plasma Heating and Current Drive.
P3T-B-300
COOLING CONCEPTS OF THE ECRH LAUNCHER STRUCTURE AND
THE TORUS WINDOWS
ROLAND HEIDINGER, IGOR DANILOV(1) GUENTHER HAILFINGER(2) KLAUS KLEEFELD(2) ANDREAS
MEIER(1) A.G.A. VERHOEVEN(3)
(1) Forschungszentrum Karlsruhe, Inst. for Materials Research, P.O.Box, 76021 Karlsruhe (2) Forschungszentrum
Karlsruhe, Inst. for Reactor Safety, P.O.Box, 76021 Karlsruhe (3) FOM Institute for Plasma Physics “Rijnhuizen”,
Nieuwegein, The Netherlands
The upper port positions for the EC wave launching system on ITER are reserved to stabilise the Neoclassical
Tearing Modes (NTM) at the q=3/2 and q=2/1 surfaces by inducing off-axis current drive. The actual mmwave system design has defined a reference beam line based on the remote steering with focusing in the
steering (poloidal) and orthogonal (toroidal) plane. The waveguide system has to be integrated into the frame
of the plug (‘main structure’) and the blanket shield module. The boundary for in-vessel components in the
port plug is set by a closure plate with CVD diamond ‘torus’ windows forming the primary tritium
confinement to the mm-wave system.The in-vessel components including the corrugated waveguides are
cooled by regular ITER blanket water from the Primary First Wall/ Blanket heat transfer system. For the
cooling of ex-vessel components, a secondary cooling system is admissible, which can be the base for cooling
of the torus windows. The cooling for the in-vessel parts is designed to provide single inlet and outlet pipe
connections with a forced sequential flow through the walls of the main structure, the blanket shield module
and the internal shields. For the waveguides an option is foreseen for lines branching off from the cooling of
the internal shield. The piping includes dog legs for thermal expansion but no double containment even outside
the closure plate. Only three joints are required for dismantling the structure by remote handling in the hot
cells. The integrated cooling concept for the launcher with details on thermal-hydraulic performance will be
presented. The CVD diamond window is exposed to non-axially symmetric thermal loads, as there is an input
steering range of up +/- 12 projected at the corrugated waveguide. Accordingly the beam center is shifted by
up to 27 mm off the window axis. The window structure is formed by copper cuffs which are brazed to the
CVD diamond disk (aperture: 95 mm) and connected to a stainless steel flange forming the outer housing.
Thermal-hydraulic and thermo-mechanical analysis was performed to show that critical stress occurs in the
OFHC copper structure. The stress levels occurring for different steering angles are discussed with respect to
their tolerance in relation to available yield strength in soft copper grades. This work is being carried out under
the EFDA technology research programme activities (Task TW3-TPHE-ECHULA and B2).
Corresponding Author:
ROLAND HEIDINGER
roland.heidinger@imf.fzk.de
Forschungszentrum Karlsruhe, Association FZK-Euratom, Institute for Materials Research I, P.O. Box 3640, D76021 Karlsruhe
71
- B - Plasma Heating and Current Drive.
P3T-B-310
THE DESIGN OF THE CONTROL SYSTEM FOR THE NEUTRAL BEAM
INJECTION IN HT-7
Z.M.LIU, XIAONING LIU SHENG LIU SHIHUA SONG DAOYE YANG YONGJUN WANG LIQUN HU CHUNDONG
HU
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031,China
The project for constructing the neutral beam injector at the Institute of Plasma Physics, Chinese Academy of
Sciences (ASIPP) was started in 2002 that is based on single injector with one arc discharge source, which can
deliver 700KW of neutral beam power at the Princeton Large Torus (PLT), USA. Initial testing during Dec.
2003 to Feb. 2004 has produced arc current up to 100 A rate for 400 msec here. The paper consists of two
parts. In the first part the distributed control system, which is the latest procedure control system that can
achieve the concentrate synthetically management in NBI are described. The second part detailed introduces
the design of each constituent part of total control system in NBI, which can complete accurate sequence
control system of the power supplies and the vacuum valves, the data acquisition and data processing. The
interlock protection system on the site is based on the programmable logical controller (PLC) system.
Corresponding Author:
Z.M.LIU
liu@ipp.ac.cn
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031,China
72
- B - Plasma Heating and Current Drive.
P3T-B-312
EXPERIMENTAL STUDY ON UNIFORMITY OF H- ION BEAM IN A
LARGE NEGATIVE ION SOURCE
HANADA MASAYA, T.SEKI, T.INOUE, T.MORISHITA, T.MIZUNO1), A.HATAYAMA1), T.IMAI, M.KASHIWAGI,
M.TANIGUCHI, K.WATANABE
Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken,311-0193, Japan 1)Faculty of Science and
Technology, Keio University, Hiyoshi, Yokohama 223-8522, Japan
The origin giving the non-uniformity of the negative ion density in the JT-60U negative ion source was
experimentally studied in the JAERI 10A negative ion source that is the same type as the JT-60U negative ion
source. Namely, the negative ion source has two different electron temperature regions divided by a transverse
magnetic field (filter field) forming uniform magnetic field along the longitudinal direction. Negative ions are
produced in the plasma with low electron temperature (<1eV) where a plasma grid (PG) is situated. The
negative ions are extracted from an ion extraction area of 15 cm x 35.6 cm. The longitudinal length is one-third
of the JT-60U negative ion source. Correlation between beam profiles and the plasma parameters such as an
electron temperature was examined. The spatial beam intensity along the longitudinal direction was relatively
low in the upper half region as observed in the JT-60U negative ion source. The electron temperature near PG
was also non-uniform even for the uniform filter filed, i.e., 1~3.5eV in the upper half region and < 1eV in the
lower half region. This high electron temperature in the upper half region was also observed in the computer
simulation. Some high-energy electrons emitted from the cathode leak to the plasma grid beyond the filter field
through the channel of a weak magnetic field of <10Gauss near the top of the negative ion source. Since the
cross-section for the destructive reaction of the H- ions via a collision with electrons rapidly increases above 1
eV, it is predicted that non-uniformity of the H- ion density in the longitudinal direction is caused by leakage
of high-energy electrons to the plasma grid. To confirm this prediction, a 50 mm x 50 mm plate for
intercepting the high-energy electrons leaked to the plasma grid was placed on the electron path predicted from
the simulation, i.e., at 7 cm from the plasma grid and 2.5 cm from the top wall of the negative ion source. The
electron temperature in the upper half region was cooled to 1 eV. This electron cooling dramatically improved
beam profile, in particularly, in the upper half region of the longitudinal direction. This resulted in a 20% gain
of the beam current. From this result, it was clarified that the leakage of the high-energy electrons to the
plasma grid is one of origins for non-uniformity of the negative ion density in the large negative ion source.
Corresponding Author:
HANADA MASAYA
hanada@naka.jaeri.go.jp
Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken,311-0193, Japan
73
- B - Plasma Heating and Current Drive.
P3T-B-314
DEVELOPMENT OF RELIABLE DIAMOND WINDOW FOR EC
LAUNCHER ON FUSION REACTORS
TAKAHASHI KOJI, S. ILLY*, A. KASUGAI, K. SAKAMOTO, R. HEIDINGER*, M. THUMM*, R. MINAMI AND
T. IMAI
Japan Atomic Energy Research Institute, 801-1, Mukoyama, Naka, Ibaraki 311-0193, JAPAN * Forschungszentrum
Karlsruhe, Postfach 3640, D-76021 Karlsruhe, Germany
A diamond window is one of important components in an electron cyclotron (EC) launcher (antenna), which
controls the injection of millimeter wave power into plasma for electron cyclotron heating and current
drive(EC H&CD). The window must have two important functions. One is the capability of high power
millimeter wave(RF) transmission. In ITER, for example, a 1MW transmission is required and has been
confirmed. Another is that to provide the vacuum and tritium barrier window, which must be the reliable
structure, between the EC launcher (vacuum vessel) and transmission lines of EC H&CD system, whose study
is reported here. When high power RF transmits through the window, it is heated up due to dielectric loss.
Therefore, the diamond window is designed to cool its disk edge to eliminate the heat deposition. Then, if it is
assumed that a crack is generated toward the window edge, for instance, by arcing or unexpected mechanical
stresses on it, the cooling medium could leak into the launcher attached to a vacuum vessel and the
transmission line. In order to avoid this possible event, the new diamond window with the copper-coated edge
has been developed. In addition, water can be used for the cooling without corrosion of aluminum blaze
between the diamond disk and the Inconel cuffs since the blaze is completely covered by Cu. To form the Cu
layer on the edge, a Ti alloy is, at first, metalized on the edge surface. Then, copper is electroformed on its
edge, entirely. The thickness of the layer is 0.5mm. A 170GHz, RF transmission experiment of the new
diamond window, which is equivalent to a MW-level transmission, was carried out to investigate that the Cu
coated window is capable of the edge cooling. The RF power of 55kW and 120kW with the pulse length up to
3sec was transmitted through the window. Temperature increases of 45 deg and 100 deg were obtained at each
RF power and they almost became constant. Thermal calculation with loss tangent of 4.4E-4 and thermal
conductivity of 1.9±0.1kW/m/K was also carried out and the result agrees with the experiment. Since the loss
tangent of the diamond used for the experiment is 4.4E-4, much higher than the actual diamond disk (loss
tangent=2.0E-5), the temperature increases correspond to those of the 1MW and 2MW transmission,
respectively. It concludes that the Cu coating on the edge dose not degrade the edge cooling capability of the
diamond window and improves the reliability of the diamond window.
Corresponding Author:
TAKAHASHI KOJI
takahash@naka.jaeri.go.jp
Japan Atomic Energy Research Institute, 801-1, Mukoyama, Naka, Ibaraki 311-0193, JAPAN
74
- B - Plasma Heating and Current Drive.
P3T-B-320
DESIGN OF HIGH POWER COAXIAL DC BREAK FOR ADITYA
TOKAMAK
MUKHERJEE APARAJITA, D.BORA (1) RAGHURAJ SINGH(1) H. M. JADAV(1) B. KADIA (1) R.A. YOGI (1)
BHATTACHARYA D.S.(2) RF GROUP (1)
(1) Institute for Plasma Research, Bhat, Gandhinagar – 382428. (2) Variable Electron Cyclotron Centre, Kolkata. (India).
ADITYA tokamak has been upgraded with the inclusion of Ion Cyclotron Resonance Heating (ICRH) system.
A 20 – 40 MHz, 200 kW ICRH system has been integrated to increase the plasma energy content. The
complete ICRH system has been indigenously designed, fabricated in-house including the RF generator. ICRH
system consists of rf generator, Tx-line, matching network (consists of stub tuner and phase shifter), vacuum
Tx-line and antenna. Outboard Fast Wave antenna is used as radiating element into the plasma. The return path
of the antenna will be directly connected to the vacuum vessel to avoid any unwanted backside radiation. DC
break in the transmission line is required to isolate the vacuum vessel from the HV power supply ground, to
which the transmitter is connected. A high power coaxial dc break is designed, fabricated and tested for wide
band frequency operation (20 MHz – 40 MHz) for blocking of dc on both inner and outer conductors. Design
of dc break, which is essentially a l/4 system for both the inner and outer conductors, is being done using
ANSOFT software. Current density is kept less than 10 Amp/cm2. Separation between the conductors is kept
in such a way so that it can withstand high voltages during mismatch. While designing, VSWR and insertion
loss are kept below 1.05 & 0.1 dB respectively for central frequency operation. Low power tests using
HP8753E VNA shows that dc break can be used from 22 MHz to 40 MHz with minimum attenuation and 1.25
(Max. at 22 MHz) VSWR. To test the performance away from center frequency, a high power test at 65 kW is
conducted at 24 MHz on test bench, which is in good agreement with low power test. In this paper, detailed
design and testing of the high power dc break will be presented.
Corresponding Author:
MUKHERJEE APARAJITA
apu@ipr.res.in
Institute for Plasma Research, Bhat, Gandhinagar – 382428
75
- B - Plasma Heating and Current Drive.
P3T-B-332
W7-X NEUTRAL-BEAM-INJECTION: TRANSMISSION, POWER-LOAD
TO THE DUCT AND INNER VESSEL AND CONSEQUENCES OF THE
STELLARATOR STRAY FIELD
N. RUST, M. KICK, E. SPETH
The new stellarator W7-X will be equipped with two ASDEX-Upgrade (AUG) like Neutral-Beam-InjectorBoxes for balanced injection. Each of them has the capacity to be equipped with 4 PINI-sized sources of 2.5
MW heating power per source. Because of the good confinement for fast ions of W7-X it is possible to use a
nearly radial injection geometry with an injection angle of 7.5 in the duct. This work will describe the result of
calculations with the neutral-beam-transmission code DENSB for W7-X. The result is not only the
transmission through the duct but also the power load on the duct and the W7-X inner vessel by the NBI. The
W7-X NBI duct has approximately the same size as AUG. Since however some W7-X coils are in direct
proximity to the duct there are some bottlenecks that limit the transmission. The total transmission for four
sources per box is 94% for a beamlet divergence of 1 . The source with the best transmission 96.7 % has nearly
no limitation by the coils. But even the source with the greatest limitations by the coils has still an acceptable
transmission of 89.7%. The narrow parts of the duct have special requirements for protection and cooling. The
Neutral-Beam also hits the inner wall of the W7-X plasma vessel. In case of no plasma the full Neutral-Beam
power will reach the inner wall. In this case the Neutral-Beams have to be switched off quickly, because
during this time the maximum power deposition on the plasma vessel by the NBI is 47 MW/m2. During the
normal plasma operation only the shine trough will hit the inner wall. The power load for the inner wall
depends strongly on the plasma density. As a conclusion the NBI pulse length is limited for low plasma
densities. For example the target modules will allow a NBI pulse length of 10 s for a central density larger than
5E19 m-3. The geometry of the inner vessel in the region of the NBI power load is quite complex. It consists
of normal wall elements. But the W7-X divertor, baffle and one port are also affected. All these in vessel
elements have to stand the NBI power load. Because W7-X has superconducting coils, the B-field is not
switched off between neutral beam pulses. The stray field of W7-X has a maximum of 300 Gauss near to the
NBI-Boxes. This is too much for the AUG like titanium getter pumps. A sufficient shielding with huge iron
masses is impossible. So the W7-X NBI may have to operate with fast cryo pumps.
Corresponding Author:
N. RUST
Norbert.Rust@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Association EURATOM-IPP, D-85748 Garching
76
- B - Plasma Heating and Current Drive.
P3T-B-337
DIAGNOSTICS OF THE CESIUM AMOUNT IN A RF NEGATIVE ION
SOURCE AND THE CORRELATION WITH THE EXTRACTED
CURRENT DENSITY
FANTZ URSEL, M. BANDYOPADHYAY, H. D. FALTER, P. FRANZEN, B. HEINEMANN, W. KRAUS, P.
MCNEELY, R. RIEDL, E. SPETH, A. TANGA, R. WILHELM
Neutral Beam Injection based on negative ion sources will be a major heating system of ITER. Besides arc
sources RF driven ion sources are promising candidates. Negative hydrogen ions (H- and D-) can be formed
by plasma volume and surface processes. Due to a short survival length of negative ions in the plasma the
surface process at the extraction grid is favoured. The optimisation of the surface process requires a high
atomic hydrogen density and a surface with a low work function for which cesium is commonly used: H + Cs > H-. One of the main tasks is to achieve a thin and homogeneous Cs coverage of the extraction grid by Cs
evaporation in the discharge volume with the boundary condition of minimising the consumption of cesium
and maintaining a constant Cs coverage at the grid. In order to quantify the amount of cesium in an RF
discharge optical emission spectroscopy is used as diagnostic tool. Suitable diagnostic lines are identified.
Their emission is observed using a line of sight parallel to the extraction grid at a distance of approximately 4
cm. Therefore, the evaluation of line emission refers to the amount of cesium in this plasma volume. The
volume density of cesium is obtained from a simplified population model using the corresponding rate
coefficients for electron impact excitation. The observation of lines from Cs ions allows an estimation of the
ion density of cesium. A method to deduce particle fluxes from the grid into the plasma is introduced as well.
Results will be presented for Cs containing hydrogen and deuterium discharges with and without additional Cs
evaporation. The influence of admixtures of rare gases on cesium will be shown. A correlation of line emission
with extracted current density of negative ions will be discussed. However, one has to keep in mind that
emission spectroscopy refers to the Cs amount in the observed plasma volume whereas the extracted current
density reflects the efficiency of the surface process. The obtained data will be used as a basis for a transport
model.
Corresponding Author:
FANTZ URSEL
fantz@ipp.mpg.de
Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr.2, D-85748 Garching, Germany
77
- B - Plasma Heating and Current Drive.
P3T-B-345
LONG PULSE OPERATION ON THE KAMABOKO III ION SOURCE.
DEIRDRE BOILSON(2), A R ELLINGBOE(2), H P L DE ESCH(1), R FAULKNER(2), , R S HEMSWORTH(1), , A
KRYLOV(1), P MASSMANN(1) AND L SVENSSON(1)
(1)Association EURATOM-CEA, CEA/DSM/DRFC, CEA-Cadarache, 13108 ST PAUL-LEZ-DURANCE (France)
(2)Association EURATOM-DCU , PRL/NCPST, Glasnevin, Dublin 13, Ireland
Development of negative ion sources is being carried out at the DRFC, Cadarache on the KAMABOKO
negative ion source in collaboration with JAERI, Japan. The target performance is to accelerate a D- beam,
with a current density of 200 A/m2 with <1 electron extracted per accelerated D- ion, at a discharge power of
<2 kW per litre of source volume, at a pressure of 0.3 Pa. For ITER a continuous ion beam must be assured for
pulse lengths of £3,600 s. Beam pulses of 1000 s have been demonstrated, but the current density at the
expected arc power and pressure was found to be to be low in comparison to the anticipated ³200 A/m2.
Accelerated currents of 320 A/cm2 have been accelerated for long pulse operation, but the transmission to the
calorimeter is only »50 % s. The loss of accelerated power is being investigated using additional electrical and
thermal diagnostics. During long pulse operation increasing the temperature of the plasma grid increases the
negative ion yield by £40%, substantially below that expected (100%). It has been shown that if the ion source
walls are kept cold (<36 C) the increase in negative ion yield with increased plasma grid temperature can be
>60% In an effort to understand the effect of Cs on the source behaviour and the reason for the low H- yield a
model of the dynamic behaviour of the Cs in the source was proposed and investigated. A cold Cs trap was
installed into the source onto which the Cs which would condense, and then the rate that Cs enters the plasma
could be controlled by increasing the temperature of the trap. Additionally, recent experiments suggest that the
Cs injected into the source is rendered unusable due to its “burial” under tungsten evaporated from the
filaments. A Cavity Ringdown Spectroscopy system (CRDS) has been installed on MANTIS which will allow
the quantitative determination of the D- (or H-) density »10 mm in front of the plasma grid. This will allow a
quantitative determination of the negative ion line density. If successful these data will be presented. This
paper will outline the aforementioned experiments and discuss the poor performance of the source in long
pulse operation.
Corresponding Author:
DEIRDRE BOILSON(2)
boilson@drfc.Cad.cea.fr
(2)Association EURATOM-DCU , PRL/NCPST, Glasnevin, Dublin 13, Ireland
78
- B - Plasma Heating and Current Drive.
P3T-B-351
DESIGN AND TEST OF A HV DEVICE FOR PROTECTION AND
POWER MODULATION OF 140 GHZ/1MW CW-GYROTRONS USED
FOR ECRH ON W7-X
BRAND PETER, BRAUNE HARALD (1) MÜLLER G. A. (2)
(1) Max-Planck-Institut für Plasmaphysik, Wendelsteinstr. 1, D-17491 Greifswald. (2) Institut für Plasmaforschung,
Universität Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart
The improvement of energetic efficiency of ECRH of fusion plasmas could be realized due to the development
of gyrotrons with beam energy recovery by a voltage depressed collector. Gyrotrons of this type for pulsed
operation were applied successfully at the W7-AS stellarator. Control of the gyrotron output power was
realized by modification of a HV-amplifier used for feeding the gun- anode of a triode type gyrotron used
before. For plasma heating by ECR in the stellarator W7-X under construction, 140 GHz gyrotrons with
depressed collector and 1MW cw output power have been developed. These gyrotrons are fed by two high
voltage sources: a high power supply for driving the electron beam and a precision low power supply for beam
acceleration. In addition a protection system with a thyratron crowbar for fast power removal in case of
gyrotron arcing is installed. The low-power high-voltage source for beam acceleration is realized by a HV
servo amplifier driving the depressing voltage, which can be modulated by feeding an adequate modulation
signal to the reference port of the servo amplifier. This new amplifier, designed for cw-operation, contains two
high voltage tetrodes working in push-pull giving an acceleration voltage swing up to 15 kVpp at a rise time of
750 V/ìs on a capacitive load of 1 nF. Furthermore the influence of the voltage noise of the main high power
supply on the acceleration voltage is suppressed by feedback control. The current of the gyrotron electron
beam is controlled by the cathode temperature. Therefore a precision ac/dc source is part of the crowbar desk.
In connection with an internal PLC (Siemens SPS) linked by Profibus optical fiber transceiver to the remote
system control, monitoring and setting of all relevant parameters is possible on the time scale of the data
aqusition of the PLC. For monitoring and control of signals up to a frequency of 100 kHz ADC- and DACfront ends linked by optical fibers have been developed. In the paper a description of the different modules of
the system is given. The results of the operation of the prototype device in conjunction with a gyrotron are
presented.
Corresponding Author:
BRAND PETER
brand@ipf.uni-stuttgart.de
Institut für Plasmaforschung, Universität Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart
79
- B - Plasma Heating and Current Drive.
P3T-B-353
DEVELOPMENT OF CW AND SHORT-PULSE CALORIMETRIC
LOADS FOR HIGH POWER MILLIMETER-WAVE BEAMS
BRUSCHI ALESSANDRO, SANTE CIRANT (1) FRANCO GANDINI (1) GIUSEPPE GITTINI (1) GUSTAVO
GRANUCCI (1) VITTORIA MELLERA (1) VALERIO MUZZINI (1) ANTONIO NARDONE (1) ALESSANDRO
SIMONETTO (1) CARLO SOZZI (1) NICOLÒ SPINICCHIA (1) GIULIANO ANGELLA (2) ENRICO SIGNORELLI
(2)
(1) Istituto di Fisica del Plasma CNR-EURATOM-ENEA, v.Cozzi 53, 20125 Milano, Italy. (2) Istituto per l'Energetica e le
Interfasi, CNR, v.Cozzi 53, 20125 Milano, Italy.
With the development of high power gyrotrons for fusion research, increased power handling of beam dumps
is required during the test phase of mm-wave systems. The design of the optics and the techniques suitable for
building a compact matched load for high vacuum operation, was developed, leading to two designs: one is
capable of 1 MW CW with proper cooling, the second is convenient for precise measurements of short pulses
(2MW, 0.1s.). Tests of the first version at 140 GHz, more than 0.5 MW and several seconds of pulse lengths
are envisaged at the ECRH plant built for the W7-X stellarator (in Greifswald), during the remote steering
antenna tests for ITER ECRH upper launcher. For both loads the spherical internal geometry is the same used
in the previous ones installed in the Frascati Tokamak Upgrade (FTU) ECRH Plant. The first CW sphere shell
is cast with a cooling pipe with inner diameter of 25 mm directly inserted in the wall thickness: it allows heat
removal with high efficiency with a water velocity of around 10 m/s. The short-pulse load has a thin copper
shell with a tube electroformed on the outside. Tube length, width and water flow rate were optimised to give a
good sensitivity for the instantaneous power and integrated energy measurements, derived by water
temperature jump and flow rate. Vacuum tests and X-ray analysis on the first cast shell showed problems of
tube adhesion to the casting, whose effects have been evaluated with thermal simulations. Problems were
solved in the second shell with a different casting procedure. The validity of the design was evaluated by
thermal and structural FEM analysis, both with uniform and realistic wall loading, obtained with analytic and
ray tracing modelling of the power distribution in the sphere interior; indication for improvements in the
cooling arrangement and power deposition were obtained. New components were designed: a cooled, vacuum
compatible vibrating mirror; a back-reflecting pre-load with a dedicated section for pumping, and a diagnostic
flange for monitoring the inner coating temperatures. The effect of the pre-load was evaluated with the same
ray tracing model used for the power distribution. New millimeter-wave measurements at low power and
heavy duty tests on coating materials show a margin for improvements in coating which could be exploited in
combination with a new mirror geometry, aiming at a higher power capability.
Corresponding Author:
BRUSCHI ALESSANDRO
bruschi@ifp.cnr.it
Istituto di Fisica del Plasma CNR-EURATOM-ENEA, v.Cozzi 53, 20125 Milano, Italy.
80
- B - Plasma Heating and Current Drive.
P3T-B-356
ITER-LIKE PAM LAUNCHER FOR TORE SUPRA’S LHCD SYSTEM
J. H. BELO (1), PH. BIBET, J. ACHARD, B. BEAUMONT, B. BERTRAND, M. CHANTANT, PH. CHAPPUIS, L.
DOCEUL, A. DUROCHER, L. GARGIULO, M. MISSIRLIAN, A. SAILLE, F. SAMAILLE, E. VILLEDIEU
(1) Centro de Fusão Nuclear, Associação Euratom-IST, Instituto Superior Técnico, 1049-001 Lisboa, Portugal
Advanced scenarios such as those envisaged for ITER require the development of a novel generation of LHCD
systems to achieve a very efficient cooling of the launcher, an essential necessity to remove the heat induced
by the neutron flux, the plasma radiated power and the RF losses. To meet these demanding goals a new and
innovative antenna based on the PAM concept (Passive-Active Multijunction) [Bibet, P., Litaudon, X.,
Moreau, D., Nucl. Fusion, vol. 35, 1213 (1995)] already proposed for ITER has been designed to be tested in
Tore Supra. It will launch 2.7 MW CW at 3.7 GHz with a power density of 25 MW/m2, radiating a power
spectrum peaked at N//=1.7 with a maximum power directivity near the electron cut-off density and with very
good coupling properties. This work has a threefold purpose. 1) To give a description of the antenna and of its
manufacturing and assembling processes: it uses eight klystrons to power sixteen TE•10-TE•30 mode
converters, each feeding its own three H-plane poloidal junction in turn connected to three E-plane bijunctions with 270 phasing, the antenna’s front part being made by linking plates of OFHC copper to
stainless-steel sheets via explosive diffusion bonding. 2) To study and optimise its RF components: the mode
converter in terms of conversion efficiency, overall SWR and balanced power splitting capabilities, the first
RF measurements of its prototype being presented; the main waveguide for an optimal transmission at the
fundamental mode TE•10 and dampening of higher modes, while avoiding reflection to the klystrons; the
bijunction length to enhance the plasma coupling; the impact of the mouth profile in the poloidal and toroidal
directions will be considered as will be the losses induced by the use of copper in the whole antenna; the
necessary measuring devices and their deployment will be defined in particular the waveguide-coaxial
transition used for measuring the S parameters; a study of the stability of the launcher to changes in the
reflection coefficients at the output ports will be undertaken to better ascertain its behaviour with varying
plasma properties. 3) To analyse its thermo-mechanical behaviour: thermal and mechanical stress analysis
taking into account the plasma radiated flux at the mouth and the RF losses; additional mechanical stresses due
to the eddy current induced in the launcher by disruptions combined with the residual toroidal magnetic field
have been computed as well.
Corresponding Author:
J. H. BELO (1)
belo@drfc.cad.cea.fr
Association Euratom-CEA, CE Cadarache,13108, St Paul lez Durance, France
81
- B - Plasma Heating and Current Drive.
P3T-B-359
LARGE CRYOSORPTION PUMP OF THE TEST STAND FOR THE
KSTAR NBI SYSTEM
IN SANG RYUL, W. S. SONG, T. S. KIM, B. H. OH
same as above
A test stand was built for developing and examining the ion sources and beam line components to be installed
in the KSTAR NBI system. The test stand is equipped with a 60 m3 vacuum chamber, an ion source, and one
set of beam line components. In the test stand, the hydrogen ion beam of maximum 2.8 MW (80 keV, 35 A)
will be produced with one ion source. Considering the ionization efficiency of 40~50%, the ion source must be
supplied with the hydrogen gas at a rate of up to 700 sccm to attain the beam current of maximum 35 A ion
beam. Moreover, the gas supply rate to the neutralizer should be at least 2000 sccm to keep the average
pressure higher than 3×10-3 mbar. In spite of such a large gas load, the chamber pressure should be low
enough not to diminish the neutral beam generated in the neutralizer. The key point in designing the vacuum
pumping system for the NBI test stand is how to evacuate the NBI chamber to the pressure of less than 10-4
mbar when the gas throughput is a few thousands sccm. The vacuum pump to fulfill such a requirement should
have a pumping speed of around 500,000 L/s. The only reasonable solution to this problem is to use an inchamber cryopump that can utilize the maximum pumping area available in the chamber. The cryo-pumping
system of the NBI test stand is composed of four cryosorption pump bodies, four G-M helium refrigerators and
four LN2 bottles of 150 L each. The main component of the pump body is a 20 K cryosorption panel cooled by
a G-M refrigerator. The cryopanel consists of 4 identical AC-coated rectangular plates of 145 mm×1000 mm
brazed to a center rod at intervals of 90 . The baffle and the lower thermal shield are cooled by liquid nitrogen.
The baffle consists of 50 chevron blades of 120 bending angle, each has a LN2 hole of 5 mm diameter along
the center axis of the blade. The chevron blades form as a whole a circular ring of 550 mm O.D and 356 mm
I.D. The liquid nitrogen level in the baffle blade is controlled by the weight and the vapor pressure of liquid
nitrogen in the bottle. The cooling down time of the cryopanel to 20 K was about 6 hours with a liquid
nitrogen consumption rate of about 35 L/hr. The maximum pumping speed of the cryosorption pump for the
hydrogen gas measured by the steady pressure method was about 90,000 L/s.
Corresponding Author:
IN SANG RYUL
srin@kaeri.re.kr
Korea Atomic Energy Research Institute, Dukjin-dong 150, Yuseong-gu, Daejeon, 305-353, Korea
82
- B - Plasma Heating and Current Drive.
P3T-B-362
DEVELOPMENT OF A RF SOURCE FOR ITER NBI: FIRST RESULTS
WITH D- OPERATION
SPETH, ECKEHART, H.D. FALTER, P. FRANZEN, B. HEINEMANN, M. BANDYOPADHYAY, U. FANTZ, W.
KRAUS, P. MCNEELY, R. RIEDL, A. TANGA, R. WILHELM
Max-Planck-Institut für Plasmaphysik, D- 85748 Garching, Germany, EURATOM-Association
ITER NBI requires among other elements a powerful beam source that delivers 40 Amperes of D- accelerated
to 1 MeV. Because of the low current densities accessible with negative ions, a large source of 1.5 x 0.6 m2
cross section is required with a net extraction area of about 0.2 m2. So far the reference design is based on an
arc discharge source, which however suffers from reduced availability and increased maintenance effort due to
the limited life of the filaments. As an alternative a RF source is being developed at IPP Garching within the
frame of an EFDA contract. The aim of this development is to demonstrate the required D- current density of
200 A/m2, accelerated to about 30 KeV and from a reduced extraction area. The target current density is
subject to the additional requirement of low source pressure (< 0.3 Pa) and low co-extracted electron fraction
(< 1). Till the end of 2003 the experiments had been restricted to hydrogen operation due to the neutron
radiation implications of deuterium. After having implemented remote operation of the BATMAN testbed,
deuterium operation started recently utilising a small extraction system of 0.007 m2. This paper reports the
first results with deuterium. After some optimisation concerning Cs operation current densities of 260 A/m2
for hydrogen and 170 A/m2 for deuterium have been achieved in the right pressure range. In both cases the
electron/ion ratio can be kept below 1 in a cesiated source. However, this requires biasing the plasma grid
against the source body with 10-20 V on the one hand and a sufficiently strong magnetic filter on the other
hand. It appears that deuterium requires a stronger filter field than hydrogen. In the present filter configuration
(external permanent magnets) the useful RF power seems to be limited, possibly due to the large source
volume filled by the magnetic field. The paper will describe the modifications to overcome this limitation by
studying different filter concepts. An interesting side effect is the fact, that the neutron production rate is about
a factor 40 lower than expected from positive ions. The paper will discuss possible reasons for this.
Corresponding Author:
SPETH, ECKEHART
speth@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, D-85748 Garching, Germany, EURATOM-Association
83
- B - Plasma Heating and Current Drive.
P3T-B-364
PERFORMANCE TEST OF THE LH ANTENNA WITH CARBON GRILL
IN JT-60U
SEKI MASAMI, MAEBARA SUNAO, MORIYAMA SHINICHI AND FUJII TSUNEYUKI
Japan Atomic Energy Institute, Naka Fusion Research Establishment 801-1 Mukoyama Naka-machi Naka-gun Ibaraki-ken
311-0193, JAPAN
Current profile control using lower hybrid (LH) wave is remarkably useful, for example, to obtain reversed
shear plasmas with higher confinement property. LH wave injection through a multijunction-type antenna (LH
antenna) has been contributing to various experiments in JT-60U during 10-year operation. This LH antenna,
however, was damaged due to excessive heat loads such as plasma bombardments and rf break downs around
its mouth. Then the injection power gradually decreased year by year. To recover the injection power, a carbon
grill is installed on the LH antenna mouth in JT-60U. It is a reason of adoption of carbon why a kind of carbon
material has high resisting power against heat load and low Z number. The carbon grill consists of a base
frame, an rf conductor and a carbon mouth. The base frame is welded with the original LH antenna made of
stainless steel. The rf conductor of thin copper plate is used to improve electrical conductivity between the
base frame and the carbon mouth. The carbon mouth is made of Graphite and/or CFC, and is held on the base
frame by 22 bolts. Therefore if the carbon mouth is damaged, it will be changed. It is possible to compare
between status of 6-Graphte type grills and that of 2-CFC type ones after experiment campaign. After
construction of the newly developed LH antenna with the carbon grill, conditioning of the LH antenna has
favorably done with and without plasmas. Through 9-day conditioning operation with plasmas, rf energy of ~5
MJ (~0.9 MW x 7.1 sec with duty cycle of 84 %) was injected into plasma. In this shot, good coupling
property was obtained such as reflection coefficient of ~5 % by controlling plasma-antenna distance. It is
found by dropping in one-turn voltage that about 60 % of plasma current of 1 MA was driven by LH injection.
Thus performance test of the LH antenna with the carbon grill is under going successfully.
Corresponding Author:
SEKI MASAMI
seki@naka.jaeri.go.jp
801-1 Mukoyama Naka-machi Naka-gun Ibaraki-ken 311-0193, JAPAN
84
- B - Plasma Heating and Current Drive.
P3T-B-371
FIRST RESULTS OF THE TORE SUPRA ITER LIKE ICRF ANTENNA
PROTOTYPE
K. VULLIEZ, S. BRÉMOND, G. AGARICI, B. BEAUMONT, G. BOSIA, B. CANTONE, J.M DELAPLANCHE, L.
DOCEUL, G. LOMBARD, L. MILLON, P. MOLLARD, E. PIGNOLY, S. POLI, B. SAOUTIC, E. VILLEDIEU, D.
VOLPE
Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 St Paul lez Durance (France)
Reliable coupling of Ion Cyclotron Range of Frequency power to plasma in high confinement ELMy regimes
is an essential target for ICRF systems. With the present ICRF systems, the fast (typically of the order of 100
ms) and big rise of the antenna loading (typically by a factor of 5) due to the ELMs results either in the
tripping of the RF generators or in reducing the coupled power if hybrid junction are used to isolate the RF
generator from the antenna, in any case reducing the mean power brought to the plasma. A new antenna RF
configuration, now know as the conjugate -T scheme matched on low impedance, was recently proposed and
adopted as reference design for ITER ICRF system [1]. The Tore Supra ITER-like ICRF antenna prototype
project was initiated in mid 2002 in Cadarache in order to get as quickly as possible some results on this new
scheme at reduced cost, as it was developed by modifying the existing ORNL Tore Supra antenna [2]. It aims
to be a tool to harvest experience and understandings in the operation of this new type of antenna, in particular
valuable for the ITER-like JET-EP antenna. The prototype antenna was first tested on test-bed, then assembled
on TS, and the first successful ICRF power coupling on plasma was obtained in February 2004. After a review
of the mechanical design, first RF results and lessons learned will be discussed in the paper, including
sensitivity of the matching due to the effects of mutual coupling between straps on this low impedance
matching scheme, load tolerance performance, antenna loading, power handling. [1]G. Bosia, ITER Joint
Central Team, Garching, Germany,High-power density ion cyclotron antennas for next step applications,
Fusion Science and Technology, 43(2003) 153-160. [2] K. Vulliez, G. Bosia, G. Agarici, B. Beaumont, S.
Bremond, P. Mollard, Tore Supra ICRH antenna prototype for next step devices, Fusion Engineering and
Design 66-68 (2003) 531-535.
Corresponding Author:
K. VULLIEZ
vulliez@drfc.cad.cea.fr
Association Euratom-CEA, CEA/DSM/DRFC, CE Cadarache, F-13108 St Paul lez Durance (France)
85
- B - Plasma Heating and Current Drive.
P3T-B-372
TORE SUPRA ITER-LIKE ANTENNA CHARACTERIZATION BY FEM
ANALYSIS
TESTONI PIETRO, GIUSEPPE BOSIA (1) PIERGIORGIO SONATO (2)
(1) CEA-DRFC/SCCP/GSAC Cadarache 13108 Saint Paul lez Durance (France) (2) Consorzio RFX, Associazione
EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)
The Ion Cyclotron (IC) technique for plasma heating and current drive is widely applied to the TORE SUPRA
experiment to develop an efficient, reliable and stationary system for the next step application like ITER. A
new four-elements (2poloidalx2 toroidal) IC array designed to launch up to 4 MW in the plasma in the
frequency range 40-60 MHz and in pulses up to for 30 s long has been installed in TORE SUPRA vacuum
vessel at the beginning of 2004. The array features with the same electric scheme adopted in the ITER
reference design, for which a significant tolerance to resistive load variations (such as those induced by ELMs
is predicted1). This paper describes the high-frequency electromagnetic (EM) analysis by 3D Finite Elements
Modeling (FEM) of the array and of its associated tuning system, consisting of a set of variable capacitive
reactances, connected in series with each element and paired poloidally in parallel. The finite element analysis
is based on a full-wave formulation of Maxwell's equations in terms of the time-harmonic electric field E
implemented in the ANSYS commercial code. The array/tuning system is first decomposed sections suitable to
establish boundary conditions and each section is accurately modeled to deduce RF electric fields and currents
distribution, so as to assess its voltage standoff capability and ohmic losses. Global (S) parameters are then
computed as function of frequency (10 to 80 MHz) for each component in standard matching conditions. The
array is finally re-synthesized by a computer code, which uses the S-parameter description to assess field and
current distributions at the appropriate tuning conditions. This method is a new approach to the design of high
frequency systems, when its physical dimensions are a non negligible fraction of the wavelength, and the
effects of local modes cannot be accounted for by a pure transversal electromagnetic mode analysis. 1) G.
Bosia “ High power density Ion Cyclotron antennas for next step applications” Fusion Science and Technology
43,153 (2003)
Corresponding Author:
TESTONI PIETRO
ptestoni@diee.unica.it
Electrical and Electronics Engineering Dept. - University of Cagliari
86
- B - Plasma Heating and Current Drive.
P3T-B-382
MAINTENANCE SCHEMES FOR THE ITER NEUTRAL BEAM
INJECTOR TEST FACILITY
ZACCARIA PIERLUIGI, A. ANTIPENKOV (3), V. ANTONI (1), A. CONIGLIO (1), S. DAL BELLO (1), C. DAY
(3), M. DREMEL (3), R. HEMSWORTH (2), T. JONES (6), A. MACK (3), D. MARCUZZI (1), A. MASIELLO (1),
M. PILLON (4), S. SANDRI (4), E. SPETH (5), A. TANGA (5), PL. MONDINO (7)
(1) CONSORZIO RFX, Padova, Italy (2) CEA, Cadarache, France (3) FZK, Karlsruhe, Germany (4) ENEA, Frascati, Italy
(5) IPP, Garching, Germany (6) UKAEA, Oxford, United Kingdom (7) EFDA CSU, Garching, Germany
The aim of the ITER Neutral Beam Injector (NBI) Test Facility is to build and test the first 16 MW NBI for
ITER and to demonstrate its reliability at the maximum operation parameters foreseen for ITER: power
delivered to the plasma 16 MW, beam energy 1 MeV, D- ion current 40A, pulse length 3600 s. ENEA-RFX
(I), CEA (F), FZK (D), IPP (D) and UKAEA (UK), are involved in an EFDA contract for the ITER NBI Test
Facility design. On the basis of the present experience on existing test facilities and of the preliminary
experimental program to be carried out on the ITER NBI Test Facility, several interventions for maintenance
and modifications are foreseen in order to optimize the beam generation and steering. The maintenance
scheme is therefore very important for the Test Facility design in order to maximize the time devoted to the
test programme. The paper describes consistently the many interrelated aspects that have been considered
during the design phase, such as: the interfaces with auxiliary systems, the need of special handling tools,
equipments and cranes, the diagnostic and monitoring systems and remote handling capabilities. Further
design requirements derived from the need of testing in advance the remote handling operations and tools
foreseen for the ITER NBI. Lifting from the top and/or running from front and rear accesses were considered
for the assembly/disassembly of the in-vessel components. Side and top access were designed, together with
equipments and fixtures that facilitate personnel access and operations in the most critical zones. Self centering
alignment systems were foreseen to speed up all the assembly and disassembly operations. The paper describes
the hydraulic, electrical, gas and mechanical connections of all the in-vessel components designed to minimize
the need of personnel access into the vessel. Optical lines of sight are located on the beam line vessel to get
optimal diagnostic and monitoring information during the operations. CCD and IR cameras will look at the
areas undergoing the most intense heat fluxes: leading edges of the neutralizer, entrance/exit of the residual ion
dump, V-shaped panels of the calorimeter. Finally the paper presents a design of cryogenic panels compatible
with the abovementioned maintenance and monitoring requirements.
Corresponding Author:
ZACCARIA PIERLUIGI
pierluigi.zaccaria@igi.pd.cnr.it
Consorzio RFX - Corso Stati Uniti,4 - 35127 Padova, Italy
87
- B - Plasma Heating and Current Drive.
P3T-B-385
NEUTRAL BEAM INJECTION OPTIMIZATION AT TJ-II
FUENTES CANDIDA, M. LINIERS (1), G. WOLFERS (1), J. ALONSO (1), G. MARCON (1), R. CARRASCO (1),
J. GUASP (1), M. ACEDO (1), E. SÁNCHEZ (1), M. MEDRANO (1), A. GARCÍA (1), J. DONCEL (1), C.
ALEJALDRE (1), C.C. TSAI (2), G. BARBER (2), D. SPARKS (2)
(1) Laboratorio Nacional de Fusión, Asociación EURATOM-CIEMAT, Av. Complutense 22, 28040 Madrid, Spain (2) Oak
Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6169, USA
The TJ-II stellarator is beginning experiments with neutral beam injection (NBI) after an experimental phase
with ECRH only. The first of two tangential injectors is now fully operative, the ion source has been
conditioned up to 30 kV accel voltage, 50 Amps extraction current. The H0 beams with power in excess of 300
kW during 200 msec, encounter a target ECH plasma of average line density 1.0 1019 m-3 and 1.5 keV
electron temperature. TJ-II is the first heliac experiment that makes use of NBI. The task is particularly
challenging in this machine because of the extremely wide magnetic axis excursion (15 % of the major radius)
and the relatively small size of the device. For this reason, optimisation of the injected beam power must be
carefully performed. Beam transmission and thermal loads in the duct, first toroidal field coil, and Circular
Coil groove inside TJ-II have been shown to depend critically on beam orientation. 3D beam simulation
studies give 50 % variations of maximum power density in the duct for a horizontal beam deviation of 0.5º.
The thermal load due to beam impact on the first toroidal field coil can reach 1100 W/cm2 for a beam tilted
0.5º in the opposite direction. Graphite protection plates have been installed at several locations inside TJ-II,
and an infrared camera surveys the hot spots along the beam from a window located on the beam duct. Beam
alignment is monitored by means of two sets of symmetrically located thermocouples. The power on the Vcalorimeter and the duct diaphragm is measured as a function of beam orientation. The reionization of the
neutral beam in the beam box and duct may account for a considerable amount of power loss. Reionization
depends strongly on the residual gas pressure in the beam box, and therefore, on the gas inventory of the
discharge. Computer simulation studies show that in order to maintain reionization losses below 10%, the
residual gas pressure must be kept below 10-4 mbar during the beam pulse. Our efforts have been aimed to
optimize gas use in the ion source and neutralizer. Fast ion gauges have been installed on the ion source and
beam box that allow us to characterize gas flow and monitor the pressure in the injector during the pulse. The
Halfa signal from a monitor located in the beam duct is related to the reionization losses. Calorimetric
measurements of beam power and neutralization fraction are compared with Halfa measurements to determine
the optimum gas injection scenario.
Corresponding Author:
FUENTES CANDIDA
candi.fuentes@ciemat.es
Laboratorio Nacional de Fusión, Asociación EURATOM-CIEMAT , Av. Complutense 22, 28040 Madrid, Spain
88
- B - Plasma Heating and Current Drive.
P3T-B-387
STATUS OF THE 140 GHZ / 10 MW CW TRANSMISSION SYSTEM FOR
ECRH ON THE STELLARATOR W7-X
KASPAREK, WALTER, H.BRAUNE(2), G.DAMMERTZ(3), V.ERCKMANN(2), G.GANTENBEIN(1),
F.HOLLMANN(2), M.GRÜNERT(1), H.KUMRIC(1), L.JONITZ(2), H.P.LAQUA(2), W.LEONHARDT(3),
G.MICHEL(2), F.NOKE(2), B.PLAUM(1), M.SCHMID(3), T.SCHULZ(2), K.SCHWÖRER(1), M.THUMM(3),
M.WEISSGERBER(2)
(1) IPF, Universität Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart, Germany. (2) MPI für Plasmaphysik, EURATOMAssociation, Greifswald, Germany. (3) FZ Karlsruhe, IHM, Association Euratom-FZK, Karlsruhe, Germany.
The stellarator W7-X which is currently under construction at IPP-Greifswald, Germany, will be equipped
with a 10 MW ECRH system, working at 140 GHz in cw regime. The microwave power will be generated by
10 gyrotrons delivering 1 MW each and will be transmitted from the gyrotron hall to the W7-X stellarator
ports via a fully optical system. The transmission system consists of 10 short single-beam mirror sections
including matching optics and polarizers for each gyrotron, and two multi-beam mirror sections (appr. 44 m)
which transmit 5 beams each to the torus hall. Near to the stellarator ports, the beams are separated again and
launched by individual antennas to the plasma. The launchers allow for arbitrary toroidal (EC-current drive)
and poloidal (on/off axis heating) launch angle of each beam. All mirrors (more than 160) are water-cooled
and can be adjusted remotely. The status of the construction of the transmission lines and the design of the
launchers is reported. Low-power tests of a prototype system at IPF Stuttgart are reviewed, showing high
transmission performance (efficiency 90 %, mode purity 98 %). The first gyrotron is operating at IPP
Greifswald, and high-power long-pulse tests have started. Measurements on transmission performance,
behaviour of the water-cooled mirrors under thermal and microwave loads as well as alignment issues,
characteristics of directional couplers, calorimetric loads and other diagnostics are discussed. The system is
presently being prepared for high-power tests of a mock-up for the remote steering antenna as planned for
ECRH/ECCD on ITER. First results of these experiments are presented.
Corresponding Author:
KASPAREK, WALTER
kasparek@ipf.uni-stuttgart.de
Institut fuer Plasmaforschung, Universitaet Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart
89
- B - Plasma Heating and Current Drive.
P3T-B-392
THE TEST OF A PAM LAUNCHER ON FTU: THE FIRST STEP
TOWARD THE LHCD LAUNCHER FOR ITER
MIRIZZI FRANCESCO, MARIA LAURA APICELLA(1), PHILIPPE BIBET(2), GIUSEPPE CALABRÒ(1), LUIGI
PANACCIONE(1), VINCENZO PERICOLI RIDOLFINI(1), SALVATORE PODDA(1), ANGELO ANTONIO
TUCCILLO(1)
(1)Associazione EURATOM ENEA sulla Fusione, C.R. Frascati, via Enrico Fermi 45, 00044 Frascati, (Rome) Italy
(2)Association CEA-EURATOM sur la Fusion, Centre d'Etude de Cadarache, France
Selected also for oral presentation
O3A-B-392
A scaled prototype of the Passive Active Multijunction (PAM) launcher actually proposed for the LHCD
system of ITER, has been realised and successfully tested on FTU in the frame of a collaboration between
ENEA Frascati and CEA Cadarache. A power density of about 80 MW/m2 at the launcher mouth
(corresponding to 50 MW/m2 in ITER at 5 GHz) has been routinely achieved with a power reflection
coefficient r ? 2% at the launcher input. A very good coupling has been obtained also with plasma density, in
front of the launcher, close to the cut-off value. Direct comparisons with the performances of a conventional
grill in a different FTU port, thus in the same operative conditions, are available. The PAM is characterised by
thick vertical walls between adjacent columns of active (transmitting) waveguides that assure a good
mechanical stiffness to the structure, while ducts drilled in these walls, allow an effective water cooling. These
thermo-mechanical characteristics make the PAM launcher very attractive for the use in the harsh plasma
environment of ITER. The periodicity of a conventional multijunction launcher is restored by interposing
columns of passive (reflecting) waveguides between the active ones at the mouth of the PAM. The directivity
of the launcher is improved at low plasma density, where the cross coupling between active and passive
waveguides is increased by strong RF power reflection conditions. This allows safe but efficient operations
with the launcher positioned far from the plasma scrape-off layer where also the thermal loads are smaller. A
full-scale test of the technological aspects of the PAM is now in preparation on Tore Supra. The proposed
LHCD launcher for ITER will couple to the plasma a total power of 20 MW at 5 GHz. The launcher is
composed of four independent units, each one including 12 PAM modules arranged in 4 poloidal rows and 3
toroidal columns. Each module is made of three rows with 8 active waveguides and 8 passive ones per row.
The intrinsic phase shift between adjacent columns of active waveguides in the same module is set to 270 to
have an N|| (peak) = 2 and an N|| = 1.9÷2.1 by changing the feeding phase between modules on the same row
in the range ±90 . The maximum RF power density at the mouth is limited to 33 MW/m2, a value that has been
largely demonstrated as safe during the test on FTU. The paper reports the main results of this test and their
fall-out on the main features of the proposed launcher for ITER.
Corresponding Author:
MIRIZZI FRANCESCO
mirizzi@frascati.enea.it
Associazione EURATOM-ENEA sulla Fusione, CR Frascati Via Enrico Fermi 45, 00044 Frascati (Rome), Italy
90
- B - Plasma Heating and Current Drive.
P3T-B-410
MATERIAL PROCESSING AND PROTOTYPE FABRICATION OF
HEAT TRANSFER ELEMENTS FOR SST-1 NBI SYSTEM.
C ROTTI, P K JAYKUMAR, K BALASUBRAMANIAN, A K CHAKRABORTY, S K MATTOO AND NBI TEAM
Institute for Plasma Research, Bhat, Gandhinagar-382 428, Gujarat, India Non-Ferrous Materials Technology Development
Centre, Kanchan bagh, Hyderabad-500 058, Andhra Pradesh, India
Heat Transfer Elements (HTEs) based upon Cu-Cr-Zr alloy are used for thermal management of the beamline
of the neutral beam injector for SST-1. It requires a thermo mechanical treatment during its fabrication which
consists of EB welding and milling. Following the recognized material production process of Cu-Cr-Zr alloy
with solution heat treatment, quenching and subsequent aging for ~ 4 hrs at 470 C has yielded mechanical
properties of the material which are in the low end of the published database. In this paper we show that
introduction of a significant percentage of cold work on the alloy yields remarkable enhancement of
mechanical properties. A cold work of 60 – 90% on solution heat treated alloy (980 C for 20 minutes) of CuCr (0.8%)-Zr(0.08%) leads to optimum properties (UTS >400 MPa, YS> 300 MPa and % elongation~22%) of
the alloy suitable for fabrication of HTE’s. Suitability for fabrication was benchmarked by detailed
characterization of EB weld joints for joint strength, micro hardness across the joint and metallographic
properties and subsequently a full scale prototype of HTE. These results shall be discussed in this paper, with a
recommendation a large database may be built for this route of material production.
Corresponding Author:
C ROTTI
crotti@ipr.res.in
NBI Group, Institute for Plasma Research, Bhat, Gandhinagar-382 428, Gujarat, India
91
- B - Plasma Heating and Current Drive.
P3T-B-412
AN ALTERNATIVE SCHEME FOR THE ITER NBI POWER SUPPLY
SYSTEM
TOIGO VANNI, A. DE LORENZI, E. GAIO, F. MILANI, L. ZANOTTO
Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova, Italy
This paper describes an alternative scheme developed for the ITER Neutral Beam Injector (NBI) Power
Supply System. The main modification proposed regards the Ion Source Power Supply (ISPS), which presents
quite high current levels and very low voltages. In the Reference Scheme, this power supply is divided in two
sections: in the first, the required power is obtained from ground referenced power regulators and then raised
to the high voltage level (-1 MV) through eight individual insulating transformers. Then the power is
transmitted via an SF6 insulated HV Transmission Line to a large tank, named High Voltage Deck (HVD),
insulated in high pressure SF6 as well. The HVD contains the final step-down transformers and diode
rectifiers, which are connected to the Ion Source by means of a second Transmission Line. The inspection of
the devices placed inside the HVD is not an easy task, due to the SF6 environment and the closeness to the
neutronic area. For this reason, and to allow easier tuning and setting up of the whole system, maintenance,
trouble shooting, fault inspections and implementation of further improvements, the possibility to eliminate the
HVD and thus guarantee full accessibility to all the ISPS devices has been analyzed. As a result, in this
alternative scheme all the ISPS components are installed inside an air insulated Faraday Cage (–1MV to
ground), the HVD is removed and the power is transmitted to the ion source via a unique SF6 insulated HV
Transmission Line. Only one main insulating transformer is required, placed upstream the whole system. An
interesting aspect of this solution is the possibility to operate for tuning the Ion Source at lower voltage and
also without the acceleration power supply, with a very easy connection between the Faraday Cage and the Ion
Source; also voltage tests and system conditioning will be greatly simplified, using a test generator directly
connected to the Faraday Cage. The main drawback of this solution is the fact that the high current conductors
are present not only in the last part of the Transmission Line, but in the whole HV line. The paper will describe
in detail the alternative power supply scheme, will discuss advantages and disadvantages with respect to the
Reference Scheme, and will present the motivations behind the design choices and their implications. In
particular, the impact on the HV Transmission Line structure and the general power supply system layout will
be presented and discussed.
WITHDRAWN
Corresponding Author:
TOIGO VANNI
vanni.toigo@igi.cnr.it
Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova, Italy
92
- B - Plasma Heating and Current Drive.
P3T-B-439
NEUTRONICS ANALYSIS OF THE ECW LAUNCHING SYSTEM IN
THE ITER UPPER PORT
A. SERIKOV, U. FISCHER(1), Y. CHEN(1), K. LANG(1), R. HEIDINGER(1), Y. LUO(2), E. STRATMANNS(1),
H. TSIGE-TAMIRAT(1)
(1)Association FZK-Euratom, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, P.O. Box 3640, D-76021
Karlsruhe, Germany. (2)Computer and Information College, Hefei University of Technology, Hefei, Anhui 230009, PR China
The design of an Electron Cyclotron Wave (ECW) launching system for the ITER port is currently under
development by a working group from various Euratom associations. The development work includes the
ECW launcher with the waveguides, the main structural components such as the port plug, shielding and
frame, and the torus window serving as vacuum closure in the waveguides. The major neutronics tasks are (i)
to assess the neutron streaming in the waveguide channels for proofing the design limit for the radiation load
to the CVD diamond window can be met and (ii) to assess and optimize the shielding of the launching system
to ensure the radiation loads to adjacent components such as the vacuum vessel and the super-conducting
magnet coils are tolerable. In addition, it must be assured that the radiation dose levels during shut down
periods are tolerable to allow maintenance personnel access to the area inside the cryostat surrounding the
ECW launcher port. This paper presents results of neutronics analyses conducted for the ECW launching
system in the ITER upper port based on the launcher design of FOM Rijnhuizen with a twisted arrangement of
8 straight waveguides. A dedicated two-step approach is used for the neutron streaming calculations with the
Monte Carlo code MCNP in the ITER 3D geometry. In the first step, a surface source is calculated at the
region of the ECW launcher front using the standard ITER plasma volume source. In the second step, the
surface source is used for calculating the neutron flux profiles along the waveguide channels by using point
detector estimators. Shielding calculations are performed with MCNP using the importance sampling
technique to assess the required dimensions of the shield material around the waveguides. The radiation dose
levels during shut-down periods are calculated on the basis of the rigorous 2-step (R2S) computational scheme
for MCNP based shut-down dose rate calculations. With all the calculations, use is made of a standard MCNP
model of ITER (20 torus sector) to which the ECW launching system was integrated. The MCNP model of the
launcher was generated from a suitably modified CATIA model by conversion into the geometry
representation of the MCNP code using a newly developed interface programme.
Corresponding Author:
A. SERIKOV
serikov@irs.fzk.de
Association FZK-Euratom, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, P.O. Box 3640, D76021 Karlsruhe, Germany
93
- B - Plasma Heating and Current Drive.
P3T-B-455
THE ITER-LIKE ICRF ANTENNA FOR JET
FREDERIC DURODIE, PH. CHAPPUIS(1), J.FANTHOME(2), R.H.GOULDING(3), J.HOSEA(4), P. U.
LAMALLE(5), A.LORENZ(6), M.NIGHTINGALE(2), L.SEMERARO(7), F. WESNER(8)
(1)CEA, (2)UKAEA, (3)ORNL, (4)PPPL, (5)LPP-ERM/KMS, (6)CSU-JET, (7)ENEA, (8)IPP-MPG
The aim of the ITER-like ICRF Antenna for JET[1] is to demonstrate novel antenna design principles in
conditions as relevant as possible to ITER in order to validate them for an ICRF heating system for ITER. The
power density for a given maximum voltage in the circuit is maximized using poloidally short straps and the
resilience to fast varying RF loads by matching pairs of straps by a so-called conjugate-T. The main challenges
during the design phase have been (i) to make the launcher - and in particular the in-vessel matching capacitors
- resilient against disruption-induced mechanical loads, and (ii) the integration of in-vessel actuators for the
matching capacitors with an overall accuracy better than 0.1mm. The antenna strap feeders were optimized to
reduce as much as possible the electrical field in the regions of highest voltage near the matching capacitors.
The tests on the High Power Prototype (ORNL and PPPL) have confirmed the voltage standoff as well as the
overall RF design but have uncovered a number of issues, such as the thermal stress in the antenna straps and
antenna box sidewalls. The results have allowed correcting the design in time for the start of the manufacturing
phases for the different components. The design phase is completed and the procurement phase of 14 packages
comprising the whole of the required hardware is now under way. Given the delivery times currently expected
for the main antenna components, the launcher is expected to be installed on the JET torus by early 2005,
following its assembly and testing on the testbed in autumn 2004. The experimental campaigns of 2005 are
proposed to take into account the challenging task of matching the 4 feed lines of the whole antenna array
simultaneously, concentrating first on delivering ICRH power for a limited number of frequencies in RFoptimized plasma conditions, and gradually broadening the operational domain towards achieving a power
density of 8 MW/m2 in ITER-relevant plasma conditions and scenarios. As part of this effort, feeding two A2
antenna arrays in a load tolerant way from 3dB couplers, similarly to ASDEX-U, will further increase the total
ICRF power available to JET. The paper will report the main characteristics of the final design as well as the
main challenges encountered during the manufacturing phase of the key components. 1. Durodié, F., et al., in
Radio Frequency Power in Plasmas, AIP Conference Proc., 595, NY: Melville, 2001, pp. 122-125.
Corresponding Author:
FREDERIC DURODIE
frederic.durodie@rma.ac.be
LPP-ERM/KMS - Avenue de la Renaissance 30, B-1000 Brussels
94
- B - Plasma Heating and Current Drive.
P3T-B-460
EFFECTS OF MUTUAL COUPLING ON ICRF LOAD-TOLERANT
ANTENNAS
SWAIN, D.(1), GOULDING, R.(1) LIN, Y.(2) PARISOT, A.(2) WUKITCH, S.(2)
1 - Oak Ridge National Laboratory 2 - Massachusetts Institute of Technology
The ability to efficiently couple radio-frequency power in the ion cyclotron range of frequencies (ICRF) can be
significantly improved, in principle, by the use of so-called load-tolerant antenna designs. The central concept
utilized is the “conjugate-tee” arrangement, in which two nearly identical current straps are symmetrically fed
through a resonant loop, producing a resistive impedance at the feed point which varies slowly as the plasma
resistance increases. Through the use of this concept, the amount of power reflected back to the transmitters
during plasma load transients can be substantially reduced. This improves the ability of the ICRF system to
continue delivering rf power in the presence of ELMs and L- to H-mode transitions. Tests of load tolerant
antenna feed configurations have been undertaken on JET, Tore Supra, and Alcator C-Mod, and a new loadtolerant antenna is presently being built for installation on JET. Initial work on load-tolerant matching was
done using relatively simple models that neglected mutual coupling between neighboring straps, but most
present and proposed future antenna designs have multiple straps, usually with significant mutual inductive
coupling. The load-tolerant properties of the uncoupled circuit can be significantly modified, and in some
cases destroyed, by the inclusion of inter-loop coupling. Initial results from a load-tolerant matching
experiment on C-Mod demonstrate that the load-tolerant characteristic can be destroyed by coupling between
antenna elements. This paper will examine the changes that mutual coupling cause on several load-tolerant
antenna concepts, and the possible application of these concepts to an antenna for ITER will be examined.
Corresponding Author:
SWAIN, D.(1)
swaindw@ornl.gov
PO Box 2008, Oak Ridge National Lab, Oak Ridge, TN 37831-6169
95
- B - Plasma Heating and Current Drive.
P3T-B-479
140-GHZ HIGH-POWER GYROTRON DEVELOPMENT FOR THE
STELLARATOR W7-X
DAMMERTZ, GUENTER, D. BARIOU(5), P. BRAND(4), H. BRAUNE(3), V. ERCKMANN(3), G.
GANTENBEIN(4), E. GIGUET (5), W. KASPAREK(4), H.P. LAQUA(3), C. LIEVIN(5), W. LEONHARDT(1), G.
MICHEL(3), G. MUELLER(4), G. NEFFE(1), B. PIOSCZYK(1), M. SCHMID(1), M. THUMM(1,2)
(1)FZK Karlsruhe,IHM,EURATOM-FZK,Postfach 3640,D-76021 Karlsruhe,Germany (2)Universitaet
Karlsruhe,IHE,Karlsruhe, Germany (3)MPI fuer Plasmaphysik,Greifswald,Germany (4)Universitaet Stuttgart,IPF,Stuttgart
Germany (5)Thales ED,Vélizy,France
Electron cyclotron resonance heating (ECRH) has proven to be one of the most attractive heating schemes for
stellarators, as it provides net current free plasma start up and heating. Extensive measurements on stellarators
at IPP Garching yield a solid physical and technological basis for ECRH systems. Therefore, ECRH was
decided to be the main heating method for the Wendelstein 7-X stellarator (W7-X) now under construction at
IPP Greifswald/Germany. A 10 MW ECRH system with continuous wave possibilities operating at 140 GHz
will be built up to meet the scientific goals of the stellarator with inherent steady-state capability at reactor
relevant plasma parameters. Two prototype gyrotrons with an output power of 1 MW were developed in
collaboration between European research laboratories and European industry (Thales Electron Devices,
France). The gyrotrons are equipped with a single-stage depressed collector (SDC), an optimized quasi-optical
mode converter and a CVD-diamond window. The prototypes have been successfully tested at FZK. With the
second one an output power of 0.89 MW with an efficiency of 42% at a pulse duration of 3 minutes and an
output power of 0.54 MW for about 15 minutes has been obtained. The first prototype has been installed at IPP
Greifswald and has been tested there successfully up to a pulse-length of 10 s. The development has been
finished; the series gyrotrons and the superconducting magnet systems have been ordered. In a parallel
development at CPI (Communications and Power Industries, Palo Alto, California) another tube had been
tested at high power (920kW, 5 ms) and delivered a power of about 500 kW with an efficiency of 35% (SDC)
in pulse-lengths of 700 s.
Corresponding Author:
DAMMERTZ, GUENTER
guenter.dammertz@ihm.fzk.de
Forschungszentrum Karlsruhe, IHM, Postfach 3640, D-76021 Karlsruhe, Germany
96
- B - Plasma Heating and Current Drive.
P3T-B-501
DEVELOPMENT OF THE 140 GHZ GYROTRON AND ITS
SUBSYSTEMS FOR ECH AND ECCD IN TEXTOR
J. SCHOLTEN(1), J.W. OOSTERBEEK(2) A.F. VAN DER GRIFT(1) J.A. HOEKZEMA (2) O.G. KRUIJT (1) A.J.
POELMAN (1) P.R. PRINS (1) E. WESTERHOF (1) C.J. TITO (1) W.A. BONGERS (1) M. R. DE BAAR (1)
Partners in TEC: (1) FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, The Netherlands,
www.rijn.nl (2) Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, EURATOM Association, D-52425, Jülich,
Germany
A 1 MW, 140 GHz gyrotron has been mounted on TEXTOR. First results on heating, current drive and
manipulation of islands have been obtained during the 2003 campaigns. Arcs in the transmission line, and the
restricted pulse length due to the limited power handling of the window and the launching mirror, hampered
the gyrotron operation. Measurements of power absorption in critical components in the quasi-optical
transmission line, in particular the quartz tokamak window and launching mirror, are presented. Experience
shows that, when the transmitted power through the plasma is high, the protection tiles for the DED on the
high field side of the tokamak can be damaged. Sniffer probes are mounted to monitor the transmitted power
levels. To avoid gyrotron operation in a wrong mode, a protection is set-up, based on a notch filter for
detecting power at frequencies outside a narrow band around 140 GHz. The following measures are being
taken to solve the various problems: A CVD diamond tokamak window will be mounted to limit absorption
and reflection from the tokamak window. Full shielding of the transmission line has been mounted. To avoid
arcs, cooling water in Teflon hoses absorbs stray radiation at critical positions in the transmission line. A new
launcher, which can couple 1 MW of EC-waves for 10 s will be installed. It features fast and accurate poloidal
and toroidal steering, for fast tracking of magnetic islands.
Corresponding Author:
J. SCHOLTEN(1)
scholten@rijnh.nl
(1) FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, P.O.Box 1207, 3430 BE
Nieuwegein, The Netherlands, www.rijn.nl
97
- B - Plasma Heating and Current Drive.
P3T-B-507
DESIGN OF CRYOSORPTION PUMPS FOR TESTBEDS OF ITER
RELEVANT NEUTRAL BEAM INJECTORS
DREMEL MATTHIAS, A. MACK(1), C. DAY(1), H.JENSEN(1), E. SPETH(2), H.D.FALTER(2), R.RIEDL(2),
JJ. CORDIER(3), B.GRAVIL(3)
(1) FZK, Karlsruhe, Germany (2) IPP, Garching, Germany (3) CEA, Cadarache, France
Special cryosorption pumps based on the adsorption with activated charcoal, coated onto stainless steel panels
are being developed at Forschungszentrum Karlsruhe in Germany. A 1:2 scaled ITER torus cryopump has
been manufactured and is under testing in the TIMO (Test Facility for ITER Model Pump) testbed. The results
of the experimental data from TIMO are used to establish and develop the design of huge cryosorption pumps,
which will be installed in the testbeds of Neutral Beam Injectors. This paper presents the results of the design
investigations and the manufacturing of two cryosorption pumps for the Neutral Beam Testbed L6 at Max
Planck Institute for Plasma Physics (IPP) and the cryopump design investigations for an EFDA contract for the
ITER NB Test Facility design. The cryopumps for IPP are foreseen to pump a hydrogen-flow of 3Pam3/s from
the beam line with a pumping speed of 350m3/s per pump. The pressure conditions must be maintained over 4
hours pumping without regeneration of the cryopanels. For cooling liquid helium at saturation pressure is used
and therefore a two-phase flow in the cryopanel system must be controlled. The hydrodynamic calculations
about the cooling with a forced flow of liquid helium are presented and the calculations of the heatloads as
well as the required mass flows for the cooling are contents of the paper. For the ITER Neutral Beam Test
Facility , design calculations to assess heatloads, pressure drops and pumping parameters of the cryopumps
based on experimental experience will be briefly described. We assess and discuss herein the different
cryogenic needs during the given operation scenarios of the Neutral Beam Injector.
Corresponding Author:
DREMEL MATTHIAS
matthias.dremel@itp.fzk.de
P.O. Box 3640 76021 Karlsruhe, Germany
98
- B - Plasma Heating and Current Drive.
P3T-B-512
STATUS OF THE NEW ECRH SYSTEM FOR ASDEX UPGRADE
LEUTERER FRITZ,
G.GRUENWALD,F.MONACO,M.MUENICH,H.SCHUETZ,F.RYTER,D.WAGNER,H.ZOHM,T.FRANKE
W.KASPAREK1),G.GANTENBEIN1),H.HAILER1),
G.DAMMERTZ2),H.HEIDINGER2),K.KOPPENBURG2),M.THUMM2),X.YANG2)
G.DENISOV3),V.NICHPORENKO3),V.MIASNIKOV3),V.ZAPEVALOV3)
Institut fuer Plasmaphysik,85741 Garching 1) Institut fuer Plasmaforschung, Universitaet Stuttgart, 70569 Stuttgart 2)
Forschungszentrum Karlsruhe, 76021 Karlsruhe 3) Institute of Applied Physics, 603600 Nizhny Novgorod, Russia
A new ECRH system with up to 4 MW/10 sec is in construction at ASDEX Upgrade.Each of the 4 depressed
collector gyrotrons can operate at different frequencies. The first one, expected in summer 2004, can work at
105 GHz (mode TE17.6) and at 140 GHz (mode TE22.8). It has asingle disc diamond window which is
resonant at both frequencies. A second gyrotron, expected later in 2004, con operate at many frequencies
within the same interval. It has atunable double disc window. In our installation we will use 4 discrete
frequencies. Since the output beam leaves the gyrotron in slightly different directions for each frequency, the
matching optics comprises a switchable pair of phase correcting mirrors for each frequency to provide a proper
Gaussian beam. The polariser mirrors are designed for the centre frequency of 122.5 GHz, but allows to set the
required polarisation for ECRH or ECCD in the whole frquency band. Thr waveguide transmission line (70 m)
has a broadband corrugation. At the torus there is another vacuum window, either single disc or tunable double
disc, which allows transmission at any polarisation. The launching mirrors, made of graphite with a copper
coating, are fast steerable in poloidal direction. A beam scan of 10 degrees in 100 msec has been achieved.
This is intended to realise a feedback controlled power deposition, in particular for experiments on suppression
of neoclassical tearing modes, where the deposition should remain within the island even when it moves due to
the Shafranov shift. The toroidal beam steering is slow, but can be changed between pulses. The first two of
these launcher units have been installed into the torus.
Corresponding Author:
LEUTERER FRITZ
leuterer@ipp.mpg.de
Max Planck Institut fuer Plasmaphysik, Postfach 1322, D-85741 Garching, Germany
99
- B - Plasma Heating and Current Drive.
P3T-B-514
IMPROVED 118 GHZ GYROTRON FOR ECRH EXPERIMENTS ON
TORE SUPRA
LIEVIN CHRISTOPHE, C. DARBOS (2), S. ALBERTI (3), A. ARNOLD (4), D. BARIOU (1), F. BOUQUEY (2), J.
CLARY (2),J.P. HOGGE (3), M. LENNHOLM (2), F. LEGRAND (1), R.MAGNE (2), M. THUMM (4)
(1)Thales Electron Devices, 2 rue Latécoère, 78141 Vélizy-Villacoublay, France (2) Association Euratom-CEA,
CEA/DSM/DRFC, CEA-Cadarache, (3) Association Euratom-CRPP,1015 Lausanne, Switzerland (4) Association EuratomFZK, Karlsruhe, Germany
A Gyrotron operating at the frequency of 118 GHz and producing 500 kW output power has been developed
thanks to a collaboration between TED (Thales Electron Devices), Association Euratom-Confédération Suisse,
Association Euratom-FZK and Association Euratom-CEA, for ECRH (Electron Cyclotron Resonance Heating)
and current drive experiments held on the tokamak Tore Supra. Tests performed on the first series gyrotron in
Cadarache [1] have shown limitations on pulse length (about 110 s), which have been explained by the
overheating of internal components of the tube, mainly due to its inadequate cooling. Moreover, the geometry
of the launcher seemed to be at the origin of spurious frequencies oscillating inside and at the output of the
gyrotron then leading to possible degraded performances. To improve the gyrotron, new studies between the
same partners have been performed and a new tube has been built by TED, mainly with a new cooling system
and a different geometry for the launcher. The factory tests up to 500 kW 5s pulses have been completed in
January 2004 and showed a significant improvement in the gyrotron conditioning, which was rather fast
compared to previous tubes.The gyrotron is now being installed at Cadarache and will be tested in long pulse
operation up to 600s. The presentation will describe the design modifications implemented on the new
gyrotron and will show main experimental results obtained, with a special focus on long pulse tests.
References [1] Very long pulse operation of the Tore Supra ECRH system, C. Darbos et al., SOFT 2002,
Helsinki.
Corresponding Author:
LIEVIN CHRISTOPHE
christophe.lievin@thales-electrondevices.com
THALES ELECTRON DEVICES, 2 rue Latécoère, 78141 Vélizy-Villacoublay, France
100
- B - Plasma Heating and Current Drive.
P3T-B-542
MATCHING TO ELMY PLASMAS IN THE ICRF DOMAIN
NOTERDAEME J-M, J-M NOTERDAEME (2) B. BEAUMONT (3) PH. LAMALLE (4) F. DURODIÉ (4) M.
NIGHTINGALE (5) I. MONAKHOV (5) THE ASDEX UPGRADE TEAM (1) THE JET-EFDA CONTRIBUTORS
AND THE TORE SUPRA TEAM (3)
(2) Gent University, EESA Department, Belgium (3) Association EURATOM-CEA, CEA-Cadarache, France. (4) Association
EURATOM - Belgian State, LPP-ERM/KMS, TEC (5) EURATOM/UKAEA Fusion Association, Abingdon, U. K.
Selected also for oral presentation
O3A-B-542
The RF generators in the ICRF domain are meant to operate into a constant, matched impedance. The coupling
of the antennas to the plasma presents a load impedance typically much lower than the generator needs for
optimal power transfer. This mismatch is overcome by a matching system that transforms the antenna
impedance to that required by the generator. However, the antenna impedance is dependent on the plasma
density in front of the antenna. This necessitates provision of a dynamic matching system, or passive load
isolation, that maintains a matched impedance despite rapid antenna impedance variations. Different
approaches can be taken with different timescales. The choice can have a substantial impact on the system
efficiency and operational reliability. These different methods are presented and discussed. For slow
variations, the mechanical change of length of a coaxial line is sufficient. For faster variations during a
discharge, capacitors are used on Tore Supra and frequency scanning on JET. Ferrite systems, utilising the
change in magnetic properties of a ferritic material with an applied magnetic field, are being developed. The
most critical area is that of very fast variations during ELMs. ASDEX Upgrade has implemented load isolation
using 3 dB couplers, which are completely passive and very reliable, maintain good current phasing, but lead
to a power loss during an ELM. This has strongly improved the performance of the ASDEX system, with up to
90 % of the installed generator power being reliably transmitted to the antenna. JET has tested frequency
variation with the addition of appropriate line components. Whereas it was shown to work in principle, the
implementation turned out to be presently not practical. JET and Tore Supra have recently tested conjugate-T
matching on plasma, the former using an external junction and the latter a new antenna again utilising internal
capacitors. These have the advantage that there is no power reduction during an ELM, but the disadvantages
that the phasing is load dependent and the matching may demand very precise setting of the components. On
JET-EP, both conjugate matching and 3 dB couplers will be implemented. The paper also indicates where
more work is still needed for the extrapolation to ITER.
Corresponding Author:
NOTERDAEME J-M
noterdaeme@ipp.mpg.de
Max-Planck Institute for Plasmaphysik, Boltzmannstraße 2, D-85748 Garching, Germany
101
- C - PLASMA ENGINEERING AND CONTROL.
P3C-C-62
OPTIMISED MODELLING OF THE TORE SUPRA TOKAMAK FOR
PLASMA EQUILIBRIUM CALCULATIONS WITH THE PROTEUS
CODE
HERTOUT PATRICK, FRANÇOIS SAINT-LAURENT (1) FERNANDA RIMINI (1) THOMAS PELLETIER (1)
(1) Euratom-CEA Association, CEA/DSM/DRFC, CEA-Cadarache, 13108 Saint Paul Lez Durance, France
Magnetic configuration calculations with plasma equilibrium finite element codes are based upon numerical
solving of Grad-Shafranov and Maxwell equations, respectively inside and outside the plasma. These plasma
equilibrium codes require a two dimensional modelling of the tokamak, assuming a global axial symmetry to
compute the poloidal magnetic field everywhere in a meridian plane of the machine. This modelling must be
carried out very carefully in the case of machines with iron core, where the return circuits violate the axial
symmetry: for example the vertical arms are assumed to be equivalent to a vertical axis cylinder with the same
total horizontal cross section, and modelled by a thin vertical rectangle in the meridian plane meshing. The
iron magnetic permeability must be also accurately determined, and the plasma facing components must be
carefully modelled, to provide consistent results in direct or reverse plasma equilibrium computations,
respectively plasma boundary determination with a given current configuration in the poloidal field coils, and
calculation of the current configuration for a given plasma boundary: in JET plasma equilibrium calculations
with the PROTEUS code, this allows a determination of the X point position with a centimetre accuracy. For a
better prediction of optimised pre-magnetization configurations and long discharges analysis with the
PROTEUS code, the modelling of the Tore Supra tokamak iron circuit and CIEL plasma facing components
has been refined: a new meshing has been built, respecting the limiter positions and the inner surfaces of the
iron return arms. The magnetic permeability has been adjusted to minimize errors between the code results and
magnetic measurements during special shots with significant horizontal or vertical field at different levels of
iron saturation. With this new modelling, magnetic configurations with field lower than 1 mT over the whole
vacuum vessel have been predicted by the PROTEUS code and successfully applied during experiments on the
Tore Supra tokamak. Plasma slow derive during long pulses in preparation to the Gigajoule discharges has
been also confirmed by PROTEUS, which proves to be a serious candidate code for ITER plasma scenario
predictions.
Corresponding Author:
HERTOUT PATRICK
patrick.hertout@cea.fr
Euratom-CEA Association, CEA/DSM/DRFC, CEA-Cadarache, 13108 Saint Paul Lez Durance, France
102
- C - Plasma Engineering and Control.
P3C-C-77
HIGH PERFORMANCE INTEGRATED PLASMA CONTROL IN DIII–D*
HUMPHREYS, D.A., R.D. DERANIAN (1), J.R. FERRON (1), R.D. JOHNSON (1), R.R. KHAYRUTDINOV (2),
R.J. LA HAYE (1), J.A. LEUER (1), B.G. PENAFLOR (1), M.L. WALKER (1), AND A.S. WELANDER (1)
(1) General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (2) TRINITI Laboratory, Troitsk, Russia
The DIII-D mission to explore the advanced tokamak (AT) regime places significant demands on the DIII-D
plasma control system (PCS) [1], including simultaneous and highly accurate regulation of plasma shape,
stored energy, density, and divertor characteristics, as well as coordinated suppression of MHD instabilities.
To satisfy the control demands of AT operation, we apply the integrated plasma control method, consisting of
construction of physics-based plasma and system response models, validation of models against operating
experiments, design of integrated controllers which operate in concert with one another as well as with
supervisory modules, simulation of control action against off-line and actual machine control platforms, and
iteration of the design-test loop to optimize. Present work describes selected new solutions which address key
control problems in DIII-D, and the approach, benefits, and progress made in integrated plasma control at
DIII-D. One element of DIII-D AT control which has been successfully addressed is the problem of high
accuracy plasma boundary control. The problem is complicated in DIII-D by the need to produce good
performance in a wide range of shapes and configurations, as well as by a uniquely constrained PF coil circuit
and power supplies routinely operated near current limits. We describe the development and implementation of
a complete predictive solution to this problem including multivariable controllers based on novel linear
nonrigid, resistive plasma models, and nonlinear algorithms to avoid saturation and windup effects. Integrated
plasma control was essential in the successful development of the DIII-D NTM control system, which has
achieved full and sustained suppression of 3/2 and 2/1 NTM modes (separately) using the 3 MW ECCD/ECH
gyrotron system to replace missing island bootstrap current. Validated models of island response to ECCD
were used to design nonlinear controllers, which vary plasma position or toroidal field to achieve alignment of
island and ECCD deposition location. We report on design and experimental use of novel algorithms using
direct feedback on the q-profile, reconstructed with a realtime Grad-Shafranov calculation [2] including MSE
measurements. [1] B.G. Penaflor, et al., Proc. of 4th IAEA Tech. Mtg. on Control and Data Acq., San Diego,
2003. [2] J.R. Ferron, et al., Nucl. Fusion 38, 1055 (1998). *Work was supported by the U.S. Department of
Energy under DE-FC02-04ER54698.
Corresponding Author:
HUMPHREYS, D.A.
humphreys@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
103
- C - Plasma Engineering and Control.
P3C-C-79
PROGRESS TOWARDS ACHIEVING PROFILE CONTROL IN THE
RECENTLY UPGRADED DIII-D PLASMA CONTROL SYSTEM*
PENAFLOR, B.G., J.R. FERRON, C.C. MAKARIOU, B.D. BRAY, D. PIGLOWSKI, AND R.D. JOHNSON
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
This paper describes the improvements being made in the capabilities of the DIII-D plasma control system
(PCS) towards achieving optimization of pressure and current profiles in advanced tokamak discharges. Key
improvements have been increased processing power and the ability to include profile diagnostic data. The
recently completed upgrade of the PCS to Linux based Intel computers connected with 2 Gigabit/s Myrinet
networking technology has been successful in achieving its goals of increasing the overall performance and
flexibility of the system. The new Intel computing system has increased processing power by a factor 30 over
the older i860 based systems. The Myrinet fiber based network has opened the doors to the inclusion of data in
real-time from DIII-D diagnostics situated in remote locations within the DIII-D research facility. The PCS
now collects 32 channels of motional Stark effect data and uses these data for real-time computation of the
safety factor (q) profile. Electron temperature and density profile data from the Thomson scattering diagnostic
and electron temperature profile data from the electron cyclotron emission diagnostic are in the midst of being
added. Addition of ion temperature and toroidal rotation profile data from the charge exchange recombination
diagnostic is planned. Feedback control from the PCS of the electron temperature at a single off-axis point has
been demonstrated using either electron cyclotron heating (ECH) or neutral beam power. This has been used to
modify current profile evolution during plasma current ramp up. Specifics of the latest improvements to the
DIII-D PCS are detailed. *Work was supported by the U.S. Department of Energy under DE-FC0204ER54698.
Corresponding Author:
PENAFLOR, B.G.
penaflor@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
104
- C - Plasma Engineering and Control.
P3C-C-103
REAL TIME CONTROL OF FULLY NON-INDUCTIVE OPERATION IN
TORE SUPRA LEADING TO 1GJ PLASMA DISCHARGES
VAN HOUTTE DIDIER, G. MARTIN, A. BÉCOULET, B. SAOUTIC AND THE TORE SUPRA TEAM
Association EURATOM-CEA, CEA-DSM-DRFC, CEA Cadarache,13108 ST PAUL-LEZ-DURANCE (France)
Selected also for oral presentation
O1B-C-103
With the goal of addressing the critical issue of the long pulse steady-state operation of next fusion devices, the
experimental programme of the Tore Supra super-conducting tokamak has been devoted in 2003 to study
simultaneously heat removal capability and particle exhaust in a steady-state fully non-inductive current drive
discharge. After a major upgrade of all the actively water-cooled in-vessel components and associated cooling
loops, partly inductively driven discharges of more than 4 minutes were rapidly obtained at multi megawatts
level in 2002. In 2003, a better understanding of several tokamak sub-system limitations, an improvment of the
plasma position within a few millimetres range, and new real time cross controls between RF Power and
various actuators built around a shared memory network, have allowed Tore Supra to access a powerful
steady-state regime with an improved safety level. In addition to the usual real time controls, two primary
feedback loops were involved: the plasma current is controlled by the lower hybrid power level and the
primary flux consumption is adjusted to zero using the main PF power amplifier voltage. As result of these
improvements, feedback controlled fully non-inductive plasma discharges have been sustained in a steadystate regime up to 6 minutes with a new world record of injected-extracted energy exceeding 1 GJ. On the
safety aspects level, a real time control of the power deposition on plasma facing components is performed
through IR monitoring of the surface temperature of the Toroidal Pumped Limiter and the supervision of water
flows and temperatures of the cooling loops. In these discharges, the injected energy is removed steadily from
the plasma facing components by high temperature pressurized water loops. The analysis of water calorimetry
and surface temperature of the actively cooled in-vessel components indicates that the thermal equilibrium is
obtained after a few seconds and shows a very satisfactory energy balance. With regard to the particle balance,
half of the injected particles are recovered in the pumps and the other half is implanted into the wall without
any indication of saturation, even after two successive 6-minute discharges.
Corresponding Author:
VAN HOUTTE DIDIER
didier.van-houtte@cea.fr
Association EURATOM-CEA, CEA-DSM-DRFC, CEA Cadarache,13108 ST PAUL-LEZ-DURANCE (France)
105
- C - Plasma Engineering and Control.
P3C-C-114
FEEDBACK CONTROL FOR PLASMA POSITION IN HL-2A
TOKAMAK
LI BO, SONG XIANMING LI LI LIU LI WANG MINGHONG FAN MINGJIE CHEN LIAOYUAN YAN QINGWEI
Southwestern Institute of Physics, P.O. Box 432,Chengdu,Sichuan,610041, P.R.China
This paper presents the horizontal plasma position feedback control system (FBCS) for HL-2A tokamak
device. It describes the hardware configuration, the program, the control algorithm, and the experimental
results of FBCS in details. It also introduces a new model of plasma resistance for computing the required
voltage waveforms of poloidal field coils for preprogrammed plasma current. HL-2A is a tokamak device with
closed divertors. It was put into operation at the end of 2002. But Divertor configuration discharges were
achieved only after the successful development and operation of FBCS. From the engineering point of view,
controlling horizontal plasma position is to control the vertical magnetic field produced by vertical magnetic
field coil (VF), i.e. to control the power supply of VF. In FBCS, a simplified proportional-differential
controller was adopted to control the power supply of VF. To guarantee the operational safety of the power
supply, some efficient measures were taken into account. It is noticed that the industrial personal computer
(IPC) used to acquire, calculate and control was programmed to be an intelligent controller. Its function was to
receive commands and control parameters from experimental management computer (EMC) through
ETHERNET. All the commands, parameters, and discharge waveforms were set or edited on EMC according
to experimental requirements. The operational interface was carefully designed so that the operator only needs
to drag and drop by using a mouse to get desired waveforms. The waveforms were calculated with a plasma
resistance model, in which plasma resistance was determined by the following two factors: 1) plasma current
stage, 2) plasma impurity and equilibrium status. For given plasma configuration, the ratios of plasma current
to VF coil current and multiple field coils current are constant. With FBCS, the plasma was confined in the
assigned area and the good repeatability of divertor configuration plasma discharges was obtained.
Corresponding Author:
LI BO
libo@swip.ac.cn
Southwestern Institute of Physics, P.O. Box 432,Chengdu,Sichuan,610041, P.R.China
106
- C - Plasma Engineering and Control.
P3C-C-149
DIII-D INTEGRATED PLASMA CONTROL TOOLS APPLIED TO NEXT
GENERATION TOKAMAKS*
LEUER, J.A,
R.D.DERANIAN(1),J.R.FERRON(1),D.A.HUMPHREYS(1),R.D.JOHNSON(1),B.G.PENAFLOR(1),M.L.
WALKER(1),A.S.WELANDER(1),D.GATES(2),J.MENARD(2),D.MUELLER(2),G.MCARDLE(3),B. WAN(4),
M.KWON(5),R.R.KHAYRUTDINOV(6), A. KAVIN(7)
(1)General Atomics,San Diego,CA 92186-5608 (2)PPPL, Princeton,NJ (3)EURATOM/UKAEA,Abingdon, UK
(4)ASIPP,Anhui,China (5)KAIST,Daejon,Republic of Korea (6)TRINITI Lab.,Troitsk, Russia (7)ITER Naka, Japan
Current and next generation fusion experiments are increasingly expected to operate in advanced tokamak
(AT) regimes and require highly integrated, complex plasma control. The DIII-D program is dedicated to the
AT mission and has developed an extensive set of modeling, simulation, and design tools for real-time control
development to enable integrated high performance regulation of plasma shape, internal profiles, fueling,
pumping, current drive and heating[1]. The highly flexible DIII-D machine allows validation of this software
suite over a wide range of plasma parameters and configurations. The system is tightly integrated with our
state-of-the-art digital Plasma Control System (PCS)[2] enabling rapid development and testing of algorithms
prior to device implementation. This paper provides an overview of this software suite and its application to
next generation tokamaks. Modeling environment elements have been used to design controllers for devices
that use, or plan to use, the DIII-D PCS, including NSTX, MAST, KSTAR and EAST. DIII-D integrated
plasma control tools have been applied to analysis and control simulation of ITER-FEAT using a
demonstration PCS. Results of applications to these devices will be presented. The software suite consists of
detailed linear/nonlinear models of plasma and system components, simulators, control design tools, and PCS
interface/testing modules. Both rigid and non-rigid linear equilibrium models of the plasma shape are used in
device simulation and controller development. A nonlinear plasma model based on DINA[3] is used to
simulate coupled plasma shape and profile evolution. System simulation is performed using a plant model
including power supplies, PF coils, passive elements, plasma models, diagnostics and data
filtering/conditioning. The simulator connects to the actual PCS hardware (or a software version of the PCS) to
perform hardware-in-the-loop tokamak/PCS simulation. The integrated plasma control suite provides a
comprehensive environment for development and testing of complex plasma control algorithms. Applications
of the suite have identified power supply characteristics and gains required to satisfy machine design
constraints. [1] D.A.Humphreys, in Proc. 20th IEEE/NPSS, San Diego, CA (2003)-[2] B.G. Penaflor, in Proc.
4th IAEA Tech Mtg on Ctrl Data Acq, San Diego, CA (2003)-[3] R.R. Khayrutdinov, J. Comp. Phys. 109
(1993) *Work supported by U.S. Dept of Energy under DE-FC02-04ER54698 & DE-AC02-76CH03073.
Corresponding Author:
LEUER, J.A
leuer@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
107
- C - Plasma Engineering and Control.
P3C-C-155
CONFIGURATION AND PERTURBATION DEPENDENCE OF THE
NEUTRAL POINT IN JET
FABIO VILLONE, V. RICCARDO (1), F. SARTORI (1), A. CENEDESE (2), B. ALPER (1), P. BEAUMONT (2)
AND CONTRIBUTORS TO THE EFDA-JET WORKPROGRAMME
(1) EURATOM/UKAEA Fusion Assoc., Culham Science Centre, Abingdon OX14 3DB, UK (2) Consorzio RFX, Assoc.
Euratom-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127, Padova, Italy
In the past, several neutral point analyses have been carried out on JT60-U [1-3] and ASDEX-U [4]. These
works have pointed out that the direction of vertical plasma movement consequent to a radiative collapse
depends on the initial vertical position of the plasma. Indeed, there exists a vertical position (called the Neutral
Point, NP) that separates a region of upwards moving configurations from a region of downwards moving
ones. If one could place the plasma at the neutral point, it will experience ideally a zero (practically very small)
vertical motion. On JET, a number of dedicated experiments have been successfully carried out in the past [5],
confirming the existence of a NP also for this device. Moreover, a modelling activity through the CREATE_L
model [6] has also justified some unexpected experimental features. This paper presents some recent
experimental results on JET aimed at extending the investigation of operation near the NP to a different plasma
perturbation: ELMs (Edge Localised Modes) due to their significance in high performance plasma operation.
The plasma configuration studied had to be changed with respect to [5], to allow an efficient plasma heating
via NBI (Neutral Beam Injection), necessary to reach H-mode and hence have ELMs. The Vertical
Stabilization System was switched off at the H_\alpha spike, triggering the subsequent vertical motion. A
dependence of the actual excitation of the unstable vertical motion on the initial vertical position was indeed
found, although no sign change (and hence no NP) was detected in the explored range. Nevertheless, the
results obtained, together with a suitable comparison with simulations obtained with the CREATE_L code,
provide extremely useful information about the correct characterization of ELMs in terms of simplified
modelling in view of plasma control. [1] Y. Nakamura, R. Yoshino, Y. Neyatani, T. Tsunematsu, M. Azumi,
N. Pomphrey, S.C. Jardin, Nucl. Fusion 36 (5) (1996) 643-656. [2] Y. Nakamura, R. Yoshino, N. Pomphrey,
S.C. Jardin, Plasma Phys. Control. Fusion 38 (1996) 1791-1804. [3] R. Yoshino, Y. Nakamura, Y. Neyatani,
Nucl. Fusion 36 (3) (1996) 295-307. [4] Y. Nakamura, G. Pautasso, O. Gruber, S.C. Jardin, Plasma Phys.
Control. Fusion 44 (2002) 1471-1481 [5] F. Villone, V. Riccardo, F. Sartori, A. Cenedese, Fusion Eng. Des.,
Vol. 66-68 (2003) 709-714 [6] R. Albanese, F. Villone, Nucl. Fusion 38 (5) (1998) 723-738
Corresponding Author:
FABIO VILLONE
villone@unicas.it
Ass. EURATOM/ENEA/CREATE, DAEIMI, Univ. di Cassino, Via Di Biasio 43, I-03043, Cassino (FR), Italy
108
- C - Plasma Engineering and Control.
P3C-C-157
DEVELOPMENT OF THE DINA-CH FULL DISCHARGE TOKAMAK
SIMULATOR
LISTER JONATHAN, V.N. DOKUKA(1) R.R. KHAYRUTDINOV(1) B.P. DUVAL J-Y. FAVEZ V.E. LUKASH(2)
J-M. MORET H. WEISEN
CRPP-EPFL, Association EURATOM-Confederation Suisse, 1015 Lausanne, Switzerland (1)TRINITI, Moscow Region,
Russia (2)RRC Kurchatov Institute, Moscow, Russia
We have continued work on developing a well benchmarked full discharge simulator based on a MatlabSimulink version of the widely known 1.5D DINA code, referred to as DINA-CH. DINA-CH had been
previously used to simulate the effects of TCV Poloidal Field voltage signals on the equilibrium modifications,
and to follow VDE’s in the highly structured vacuum field pattern of TCV. Work has progressed on simulating
the effect of very localised Electron Cyclotron Heating and Current Drive in TCV. We have now reconciled
the different versions of DINA developed for ITER, TCV and MAST into a single version which includes the
possibility of defining plasma transport in external blocks. External heating and current drive can now be
specified both in [R,Z] coordinates, developed for ECH and ECCD on TCV, but also in flux-surface
coordinates more suitable for NBI, alpha-heating and radiation losses. We can now save the simulation data as
MDS+ data structures, allowing more effective remote access to the results using the standard interfaces to
MDS+ data. Work is currently underway to incorporate detailed up to date transport models and results should
be presented. We are presently incorporating an equilibrium to diagnostics mapping toolbox which was
originally developed for the TCV tokamak and which has been exhaustively tested and refined on the large
variety of TCV plasma shapes. An example of real-time simulation of a representative diagnostic is presented
for the case of ITER using this toolbox. The nominal ITER magnetic diagnostics are now also included in the
simulations of ITER. We are integrating an estimate of the AC losses during operation of superconducting
Poloidal Field coil magnets, previously developed and benchmarked against a full evaluation code as an ITER
design task. The estimate of the deposited heat is available during the simulation, rather than estimated after
the pulse. A full ITER simulation with all these recently implemented features will be presented to illustrate
the advances described.
Corresponding Author:
LISTER JONATHAN
Jo.Lister@epfl.ch
CRPP-EPFL, 1015 Lausanne, Switzerland
109
- C - Plasma Engineering and Control.
P3C-C-161
DESIGN, IMPLEMENTATION AND TEST OF THE EXTREME SHAPE
CONTROLLER (XSC) IN JET
ALBANESE RAFFAELE, G. AMBROSINO(1) M. ARIOLA(1) A. CENEDESE(2) F. CRISANTI(3) G. DE
TOMMASI(1) M. MATTEI F. PICCOLO(4) A. PIRONTI(1) F. SARTORI(4) F. VILLONE(5) AND JET-EFDA
CONTRIBUTORS
(1)Ass. Euratom-ENEA-CREATE, Univ. Napoli Federico II, Italy (2)Consorzio RFX, Ass. Euratom-ENEA sulla Fus., Italy
(3)Ass. EURATOM-ENEA sulla Fus., Frascati, Italy (4)Euratom/UKAEA Fusion Assoc., UK (5)Ass. Euratom-ENEACREATE, Univ. Cassino, Italy
Since in ITER the reference scenarios are planned to work at extreme plasma shape, JET operation will be
progressively focused on the study of this kind of plasmas. The old JET Shape Controller (SC) can only
control a few plasma-wall gaps at the same time. This, for strongly shaped plasmas, can lead to large
deformations of the shape, mainly in case of large variations of poloidal beta bp and/or internal inductance li.
A new JET eXtreme Plasma Shaping controller (XSC) was designed to achieve extremely shaped plasmas
with the existing active circuits and control hardware. A linearized plasma model approach was used in
designing the XSC. This allowed linking the applied active coils with any of the geometrical descriptors
characterising the plasma boundary. Several (up to 36) gaps were used to describe the plasma shape
accurately. However only a limited set of actuators is available (only 8 poloidal circuits in JET). The problem
was tackled by using a singular value decomposition (SVD) to identify the principal directions of the algebraic
mapping between coil currents and geometrical descriptors. These principal directions are then assumed as
controller inputs/outputs so that the original multivariable control problem can be solved using a set of
separate PID controllers. To take in account the limits of the actuators, the SVD orders the principal directions
as a function of the current to shape sensitivity and the XSC normally uses only the first 5 or 6 directions (out
of 8). This new system was successfully installed on the JET machine during 2003 without causing any
interference to the plasma operation, or requiring long commissioning time. Eventually the new controller was
used on really extremely shaped Internal Transport Barrier (ITB) experiments at high poloidal beta and in the
presence of quite large variations of bp (Dbp up to 1.5) and/or li (Dli up to 0.5). The quality of the model based
controller design approach was also been verified by a large sequence of plasma scenarios where extremely
elongated shapes were achieved for the first time by using the controller without requiring any kind of tuning.
The XSC controller architecture and philosophy also offer new interesting opportunities, e.g., the separatrix
sweeping on the divertor plates without significantly affecting the overall plasma shape, and the possibility of
improving the overall tokamak performance via combined control of plasma shape, current and profile.
Corresponding Author:
ALBANESE RAFFAELE
albanese@unirc.it
Assoc. Euratom-ENEA-CREATE, DIMET, Univ. Mediterranea RC, Loc. Feo di Vito, I-89060, RC, Italy
110
- C - Plasma Engineering and Control.
P3C-C-190
CORRECTION POSSIBILITIES OF MAGNETIC FIELD ERRORS IN
WENDELSTEIN 7-X
KIßLINGER JOHANN, ANDREEVA TAMARA
Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald, Germany
WENDELSTEIN 7-X (W7-X) is an advanced stellarator with an optimised magnetic field configuration with
respect of plasma confinement and stability. The super-conducting magnet system of W7-X has a modular
structure and consists of five identical field periods. In order to reach the high quality of the magnetic field and
to preserve the symmetry of the machine the magnet system need to be constructed and assembled with very
high precision. Already small statistic deviations from ideal coil shapes or small not symmetric misalignments
of the coils cause error field components which lead to additional magnetic islands and asymmetric thermal
loads on the divertor targets. The requirement of good plasma confinement and reliable divertor operation
limits the relative amplitude of error field component to be below 2 10-4. In order to reach this ambitious goal,
correction possibilities are foreseen during the assembly and, if required, by the help of correction coils.
Fabrication and assembly of the machine components are continuously accompanied with a precise
geometrical survey. The perturbing impact of measured deviations from the ideal geometry is then analysed by
magnetic field calculations and becomes the basis for adjustments. Thus, by a combination of distinct shifts
and inclinations of some coils or modules it is possible to reduce or to cancel out perturbing field components.
A numerical optimisation code was written to predict the best alignment combination of the coils on the basis
of the geometrical analysis of the manufactured coils. Despite all measurements are done with high precision,
some uncertainties of the geometrical shape of the magnet system and also adjusting tolerances will remain. In
the case, that a significant field perturbation exist in the magnetic field, correction coils are required. The 10
control coils in W7-X, planed for island shaping and sweeping, are also usable for field error correction and
are able to generate low order Fourier components up to 0.4mT. If larger correction fields are needed
additional coils must be provided. As examples for such coils the capability and technical realisation
possibility of normal-conducting coils inside the vacuum vessel, super-conducting helical windings within the
cryostat and normal-conducting saddle shaped coils on top of the outer shell are investigated.
Corresponding Author:
KIßLINGER JOHANN
johann.kisslinger@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, D-85748, Garching, Germany
111
- C - Plasma Engineering and Control.
P3C-C-207
REAL TIME CONTROL ENVIRONMENT IN THE RFX EXPERIMENT
LUCHETTA ADRIANO, BARANA OLIVIERO, CAVINATO MARIO, MANDUCHI GABRIELE, TALIERCIO CESARE
A complex, distributed, digital system has been implemented to realize a comprehensive set of control
schemes that can be applied to the RFX experiment by means of the recently enhanced power supply system
and the load assembly. Two basic feedback control schemes, the equilibrium control and the Resistive Wall
Mode stabilisation, aim at extending the pulse duration beyond the time constant of the conducting shell for
the penetration of radial field (50 ms), while other control schemes, such as the MHD Mode Control and the
‘Intelligent Shell’ aim at achieving a better understanding of the underlying physics. The distributed digital
system consists of a set of seven computing nodes where each of which can be either a pre-processing station,
in charge of acquiring raw data and processing intermediate control parameters, or a control station, in charge
of driving the actuating power amplifiers. The paper provides an overview of the implemented system
architecture with particular reference to the control schemes realized. Emphasis is given in the paper to the
software real-time environment, which is providing not only basic functions, e.g.: data read-out, real-time
communication and data collection, but also useful tools for programming and integrating control algorithms,
performing simulating control scenarios and finally commissioning the system. Time-critical algorithms are
coded directly in C fully exploiting the specific processor architecture only where necessary. This approach
produces code that runs faster but is hardly portable among processor architectures. Non time-critical control
algorithms are developed under Matlab and executed on the real-time processor. The advantage is that the
coding process needs no software specialists and control engineers can focus on the algorithms. Portability of
code depends in this case on the availability of a Matlab real-time toolbox for the candidate real-time operating
systems. To perform simulations of various control schemes the real-time software environment is being
extended to include a so called “simulation mode”. In this mode, the real-time controllers exchange their
input/output signals not with the real system, but with a station running a suitable model of the system, for
instance MAXFEA in the case of equilibrium control. In this way the control algorithm can be tested offline
and the time needed for commissioning of algorithms can be reduced.
Corresponding Author:
LUCHETTA ADRIANO
adriano.luchetta@igi.cnr.it
corso Stati Uniti, 4 35127 Padova - Italy
112
- C - Plasma Engineering and Control.
P3C-C-233
A FAST AND VERSATILE INTERLOCK SYSTEM
MICHEL, GEORG,
In todays large fusion experiments it is a common task to shut down a component quickly under certain
conditions. These conditions can imply a complex logic and timing as well as topologically distributed inputs.
In addition, the shutdown should be as instantaneous as possible. The shutdown mechanism should be flexible
enough to allow for the changing requirements in the daily experimental work. The paper presents an interlock
system which meets these reqirements. The system has a modular design consisting of an arbitrary number of
identical distributed units which are connected to the general purpose computer network and to a dedicated
interlock bus. The units consist of an FPGA (Field Programmable Gate Array) which implements the fast logic
(nanosecond scale) and a microcontroller which is connected to the FPGA via its data and address bus. The
microcontroller implements the network connectivity for the FPGA providing configuration, identification and
reset functionality (millisecond scale). The units observe analog signals for the overstepping or understepping
of a programmable threshold in a programmable time interval. Their main component is a commercial
embedded microcomputer engine with FPGA (DragonEngine II) running µClinux. The interlock bus transmits
the actual interlock and timing signals with the mere transmission line delay while the computer network is
used for all other information. Due to this combination the interlock system can be made fast and versatile at
the same time. A prototype of the interlock system has been built at IPP Greifswald and tested at the ECRH
plant of W7-X. It consists of four units with ten inputs each.
Corresponding Author:
MICHEL, GEORG
michel@ipp.mpg.de
MPI fuer Plasmaphysik, Wendelsteinstr. 1, 17491 Greifswald, Germany
113
- C - Plasma Engineering and Control.
P3C-C-268
NEW VISUALIZATION SYSTEM FOR CONTROLLING AND
MONITORING PURPOSES IN THE TJ-II STELLARATOR
LUIS PACIOS, ANGEL DE LA PEÑA, RICARDO CARRASCO, FERNANDO LAPAYESE.
The Spanish Stellarator TJ-II is a highly flexible medium size fusion device of the heliac type in operation
since the end of 1997. From its inception the TJ-II control system has performed satisfactorily over more than
10,000 plasma shots. The hardware and software, chosen ten years ago, for the control system architecture is
based on a distributed network of embedded computers in VME crates running the OS9 real time operating
system. For the realisation of the graphical user interface (GUI) of each sub-system the proprietary software
named G-Windows and a dedicated video card were employed. In accordance with the experience acquired
during the last 6 years of its operation, and bearing in mind new technological trends, the control system has
been upgraded. For this purpose web-based software tools have been applied. The new visualization system
based on Java applications frees us from the dependence on G-Windows. In addition, it provides new GUI
features such as audio and video capabilities. Systems such as gas puffing, trigger programming, ground loop
supervision, neutral beam injection control, and diagnostics can be fully configured and monitored in real time
using any web browser or Java programs. Interactions between systems are based on Client-Server architecture
and remote method invocation protocols. Two dedicated sockets in each system provide data and control flow
between the control application and either client browser or java program. This paper will describe the new
visualization system and shows the way it has been applied to real-time control applications.
Corresponding Author:
LUIS PACIOS
l.pacios@ciemat.es
Asociacion EURATOM-CIEMAT para Fusion, Avda. Complutense 22, E-28040 Madrid, Spain
114
- C - Plasma Engineering and Control.
P3C-C-277
WEB-BASED GROUND LOOP SUPERVISION SYSTEM FOR THE TJ-II
STELLARATOR
ANGEL DE LA PEÑA, FERNANDO LAPAYESE LUIS PACIOS RICARDO CARRASCO
To minimize electromagnetic interferences in diagnostic and control signals and to guarantee safe operation of
TJ-II, ground loops must be avoided. In order to meet this goal the whole grounding system of the TJ-II was
split into multiple single branches that are connected at a single grounding point situated near the TJ-II
structure in the torus hall. Each cabling rack and all the main components and structure must be grounded to
the main grounding bar via separate ground wires. A real-time Ground Loop Supervision System (GLSS) has
been designed, manufactured and tested by the TJ-II control group for detecting unintentional short circuits
between separately grounded parts. The system measures loop impedance without breaking grounding
connections in order to identify the short circuit. In accordance with TJ-II control system requirements, the
GLSS hardware selected consists of VME analog and digital I/O boards, sets of emitter and sensor coils
embracing the ground wires to be controlled and modular signal conditioning equipment. The measurement
technique is based on the current induction method already employed in other fusion laboratories, e.g. in
CRPP, Lausanne. For this a pulsed 1700 Hz sinusoidal current is continuously induced by a toroidal emitter
coil in the ground wire for periods of 40 ms. Hence, if a ground loop is present this induced current can be
detected by a current probe. With this set-up, the GLSS measurement range is between 10 ohm and 10 Kohm.
At present, eight ground branches are controlled but this number could be increased easily if needed. The
GLSS Software was developed using client-server architecture. A web server running on the Real Time
Operating System OS-9 provides the interface with application control. Java graphical user interfaces allow
remote access to the real-time ground loop measurement. Multiple mode operations and alarm thresholds can
be configured via any web browser. This paper gives the detailed design of the whole TJ-II ground loop
supervision system and its results during operation.
Corresponding Author:
ANGEL DE LA PEÑA
a.delapena@ciemat.es
Laboratorio Nacional de Fusión, Asociación Euratom-Ciemat, Avda Complutense 22, 28040 Madrid (Spain)
115
- C - Plasma Engineering and Control.
P3C-C-291
A NEW CONTROLLER FOR THE JET ERROR FIELD CORRECTION
COILS
LORIS ZANOTTO, MARCO BIGI(1), FABIO PICCOLO(1), FILIPPO SARTORI(1), MASSIMO DE BENEDETTI(2)
(1) EURATOM/UKAEA Fusion Ass. Culham Science Centre, Abingdon, OX14 3EA, United Kingdom (2) Associazione
Euratom-ENEA sulla Fusione, Frascati, Italy
Experiments about the stability of Resistive Wall Mode, error field induced locked modes and (neo-classical)
tearing modes are routinely run on JET. In some of these experiments the MHD instability is excited by means
of an external error field. In the past, such studies were performed by means of the Disruption Feedback
Amplifier System (DFAS), which was devoted to supply a set of four Saddle Coils placed inside the vessel;
the Saddle Coils are going to be completely dismissed during the 2004 shutdown. In order to continue
experiments on error fields and tearing modes, a new system of Error Field Correction Coils has been recently
installed outside the vessel aiming at producing a quasi-static error field. This new coils are supplied by the
same DFAS. In order to optimise the performances of the DFAS with the new load and to improve the
flexibility of the system, a completely new controller has been designed and implemented using a VME
hardware platform. Its main tasks are: - to provide references for the DFAS up to 10 kHz - to provide a flexible
trigger logic system - to provide reference components for the compensation of the natural error field - to
provide the possibility to perform Resistive Wall Mode stabilisation and excitation. The hardware part has
been simplified with respect to the previous solution, based on a cluster of C-40 Digital Signal Processors. All
analogue input signals are processed by a conditioning stage and then acquired by the VME crate through a
data acquisition board. Furthermore, an Asynchronous Transfer Mode interface board provides inputs from the
Real Time Data Network (RTDN) of JET. The software part of the controller is based on a 100 kHz interruptdriven algorithm and features a flexible trigger logic interface based on strings including RTDN signals,
logical and mathematical operators. Therefore, reference to the amplifiers can be enabled and disabled using
signals from the magnetic sensors or disruption precursor. Thanks to these new features and to the possibility
to easily upgrade the system, due to the modularity of the software structure, the new controller will be a
useful tool to perform error field and tearing modes studies on JET in the future. This paper will present the
hardware and software structure of the new controller, along with the new possibilities offered by its very
flexible trigger logic structure. Some results of the commissioning and the first operational experience will
also be reported.
Corresponding Author:
LORIS ZANOTTO
loris.zanotto@igi.cnr.it
Consorzio RFX - Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti, 4 35127 Padova (Italy)
116
- C - Plasma Engineering and Control.
P3C-C-298
USING REAL TIME WORKSHOP FOR RAPID AND RELIABLE
CONTROL IMPLEMENTATION IN THE FRASCATI TOKAMAK
UPGRADE FEEDBACK CONTROL SYSTEM RUNNING UNDER RTAILINUX
VITALE VINCENZO, CRISTINA CENTIOLI (1) FRANCO IANNONE (1) MAURIZIO PANELLA (1) LUIGI
PANGIONE (2) SALVATORE PODDA (1) LUCA ZACCARIAN (2)
(1) Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy (2)
Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Via del Politecnico 1 - 00133 Roma,
Italy
The feedback control system running at FTU has been recently ported from a commercial platform (based on
LynxOS) into an open-source Linux-based RTAI-LXRT platform, thereby obtaining significant performance
and cost improvements. Based on the new open-source platform, it is now possible to experiment novel control
strategies aimed at improving the robustness and accuracy of the feedback control task. Nevertheless, the
experimentation of control ideas still requires a great deal of coding of the control algorithms that, if carried
out manually, may be prone to coding errors, therefore time consuming both in the development phase and in
the subsequent validation tests consisting of dedicated experiments carried out on FTU. In this paper we report
on recent developments based on the Mathworks’ Simulink and Real Time Workshop (RTW) packages to
obtain a user friendly environment where the real-time code required to implement novel control algorithms
can be easily generated, tested and validated. Thanks to this new tool, the control designer only needs to
specify the block diagram of the control task (namely, a high level and functional description of the novel
algorithm under consideration) and the corresponding real-time code generation and testing is completely
automated without any need of dedicated experiments. In the paper, the necessary work carried out to adapt the
Real Time Workshop (RTW) to our RTAI-LXRT context will be illustrated. A necessary re-organization of
the previous real-time software, aimed to incorporating the code coming from the adapted RTW, will also be
discussed. Moreover, we will report on a performance comparison between the code obtained using the
automated RTW-based procedure and the hand-written C code, appropriately optimized: at the moment, a
preliminary performance comparison consisting of dummy non-trivial algorithms has shown that the code
automatically generated from RTW is faster (about 30% up) than the manually written one. This preliminary
result, combined with the fact that the use of RTW eliminates the coding workload leads to the conclusion that
this approach to control implementation grants significant. According to the FTU experiments program, the
actual deployment of the new tool is foreseen to be scheduled in the next experimental campaign.
Corresponding Author:
VITALE VINCENZO
vitale@frascati.enea.it
Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy
117
- C - Plasma Engineering and Control.
P3C-C-301
THE SYSTEM ARCHITECTURE OF THE NEW JET SHAPE
CONTROLLER
FILIPPO SARTORI, G.AMBROSINO(1) MARCO ARIOLA(1) GIANMARIA DE TOMMASI(1) ALFREDO
PIRONTI(1) ANGELO CENEDESE(2) FLAVIO CRISANTI(3) PAUL MC CULLEN(4) FABIO PICCOLO(4)
(1)EURATOM-ENEA-CREATE, V. Claudio 21, 80125 Napoli, Italy (2)Consorzio RFX, EURATOM-ENEA, C.so Stati Uniti 4,
I-35127 PD, Italy (3)EURATOM-ENEA, Frascati, C.P. 65, 00044-Frascati, Italy (4)EURATOM-UKAEA Fusion Ass.,Culham
Science Centre,OX143DB,UK
The DSP based JET plasma Shape Controller (SC) has been replaced by a Power-PC system. The trigger for
the work was the desire to experimentally validate the results of the Extreme Shape Controller (XSC)
enhancement. The actual implementation went much further, producing an advanced and comprehensive
solution that allows in the same pulse the free mixing of the traditional gap and current based control schemes
with the full boundary control introduced by XSC. Changes to both the software architecture of the real-time
controller and to the user interface were introduced, while still maintaining backward compatibility. The XSC
control scheme was first adapted to JET operational needs in order to allow the imposition of current limits,
and then generalised into a two-loop MIMO. The inner loop, SC, takes care of the de-coupling current control
problem. The outer, XSC, is the multi-variable shape controller. 16 inputs and 8 outputs are reserved for
external system connections to allow integration with the JET real-time control, feature that was successfully
employed in the TAE experiment. The user interface was enhanced to allow, in different time windows, the
programming of completely different control scenarios, both in term of matrices and set points. The XSC
scenarios are just one subset of these, but can also be visually inspected and interactively tuned by the final
user in order to achieve the desired objective. A major design objective was the minimisation of the
commissioning effort of the final product. Since the most complex and safety critical part of the controller is
the current and voltage limit software, the choice of interfacing XSC to SC via the existing limit avoidance
logic, allowed the minimisation of the testing time. Once the original control and protection algorithms had
been re-commissioned, it was then possible to rely on them while testing the most advanced schemes. The
combination of this system architecture, with a model based design technique allowed a speedy introduction of
the new controller, which has since been successfully operating catering for all the requirements of the
experimental program. The new JET-EP divertor design will require further evolution within the current
architecture. The sweeping logic will need to able to follow different sweeping trajectories. The termination
scenarios will have to be made more flexible and a scheme will have to be developed in order to reduce the
XSC saturation problems.
Corresponding Author:
FILIPPO SARTORI
fisa@jet.uk
EURATOM-UKAEA Fusion Ass., Culham Science Centre, OX14 3DB, UK
118
- C - Plasma Engineering and Control.
P3C-C-350
AN INTEGRAL APPROACH TO PLASMA SHAPE CONTROL
BEGHI ALESSANDRO, M.CAVINATO(2), A.CENEDESE(2), D.CISCATO(1), S.SIMIONATO(1)
(1)D.E.I. and Centro Ricerche Fusione, Università di Padova, Via Gradenigo, 6/A, I-35131 Padova, Italy. (2)Consorzio RFX
Associazione EURATOM ENEA sulla Fusione, Corso Stati Uniti, 4, I-35127 Padov
In the future generation of Tokamaks the reactor performances are strictly related to the capability of
sustaining and controlling strongly shaped plasmas. Currently the shape control is based on a pointwise
description of the plasma boundary usually in terms of a set of plasma-wall distances (gaps). In this paper an
alternative approach is presented, based on an integral description of the plasma boundary using the parametric
curve model given by the B-Splines [1]. This model permits to construct a piecewise polynomial curve globally Cn-2 if n is the B-Spline order - by linearly combining piecewise polynomial basis functions. With
such a tool it is possible to describe a wide range of curves with a relatively low number of parameters, giving
a compact characterization of the whole shape. The procedure consists in the extraction of the plasma
boundary curve from the flux map, which is then fitted by means of planar B-Splines. Two approaches to the
extraction of the boundary curve have been tested. The first is based on the active contour approach and
involves the deformation of the curve according to a functional minimisation, while the second operates
directly the fitting of the boundary curve with the parametric model. The attention was then focused on the
second approach since it offers an easier way to obtain the surjectivity in the relation shape-parameters and
presents a reduced computational cost. In order to ease the control design we studied two techniques to convert
the two-dimensional description of the curve in a one-dimensional signal with no or little loss of information.
These techniques permit to almost halve the number of integral descriptors while maintaining the same
precision. With reference to the ITER machine, the number of shape parameters was reduced to ten allowing
the design of a controller based on the decoupling technique similar to the gap controller. The comparison of
the performances given by the two control approaches is presently underway and focuses on the control of
global shape parameters such as triangularity and elongation. The integral approach is expected to give a
smoother control of these quantities which have a strong relationship to the overall plasma performances. [1]
A. Blake and M. Isard, “Active Contours” Springer, 1998.
Corresponding Author:
BEGHI ALESSANDRO
beghi@dei.unipd.it
D.E.I. and Centro Ricerche Fusione, Università di Padova, Via Gradenigo, 6/A, I-35131 Padova, Italy
119
- C - Plasma Engineering and Control.
P3C-C-352
LINEARIZED MODELS OF THE PLASMA RESPONSE IN THE NEW
RFX LOAD ASSEMBLY
BETTINI PAOLO, (1)M. CAVINATO, (1)G. MARCHIORI, (2)F. VILLONE
(1) Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova, Italy (2) Ass.
EURATOM/ENEA/CREATE, DAEIMI, Università di Cassino, Via Di Biasio 43, I-03043, Cassino (FR), Italy
The RFX load assembly now includes a new 3 mm thick copper shell, close fitting to the vacuum vessel,
which will allow to perform active control experiments by means of a set of 192 saddle coils surrounding the
torus. The much shorter time constant of the shell also requires an active control system of the plasma
equilibrium. At present, the design of the control system in fusion devices is generally based on linearized
models of the plasma response at different equilibria. In the past, the CREATE_L plasma response model,
derived by linearizing the MHD equilibrium equation and Ohm’s law in the active and passive conductors and
in the plasma, was successfully applied to a RFP plasma in the presence of the old RFX magnetic
configuration. One of the aims of this paper is to apply this procedure also to study in details the effects of the
structural modifications of RFX on the open loop plasma response. Since experimental data are not yet
available, a two dimensional FEM code, solving the ideal MHD free boundary problem in axisymmetric
geometry, has been used to provide a set of equilibrium reference data, also in terms of coil currents and
virtual measurements from pick-up and flux loop probes. As a first step in the validation procedure, in
particular to assess the accuracy of the electromagnetic model of the passive structures implemented in the
codes, a cross-check has been carried out on a shot without plasma, programmed to apply a vertical field step.
The agreement between linear and non-linear model is very good. As a second step, complete comparisons
have been performed in the case of pulses with plasma. The currents in the active coils were imposed, while
the time behaviour of the currents in the passive structures was self-consistently calculated. This was possible
thanks to the fact that the plasma is open-loop stable allowing us to focus only on the error introduced by the
open loop model. A further step in view of plasma control has been the inclusion of active coil voltages as
inputs to the model. The preliminary results are encouraging in both cases. Cross-comparisons with another
available linearized model of the plasma response, derived according to a different perturbation method, are
also presented.
Corresponding Author:
BETTINI PAOLO
bettini@uniud.it
DIEGM, Università di Udine, Via delle Scienze, 208, I-33100 Udine, Italy
120
- C - Plasma Engineering and Control.
P3C-C-354
DESIGN OF THE NEW RFX EQUILIBRIUM ACTIVE CONTROL
SYSTEM
CAVINATO MARIO, G.MARCHIORI(1)
(1)Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione,Corso Stati Uniti 4, I-35127 Padova,
The RFX new load assembly is characterized by a thin copper shell with a 50 ms time constant for the
penetration of the vertical magnetic field, shorter than the pulse length. Thus an active control system is
necessary to assure the plasma horizontal equilibrium. The design has been carried out by developing for the
RFP the classical method used in the case of ITER or for recent upgrades of presently operating Tokamak
devices. First of all the geometry of the RFX new load assembly was implemented in a FE MHD plasma
equilibrium code (MAXFEA), then equilibria at different plasma currents (750 kA and 2 MA) were calculated
and corresponding linearized models of the plasma response were derived by means of a perturbation method.
Finally, the design of the controllers was accomplished on the basis of the available linear models. A nested
loop structure was kept, following the scheme already adopted in the old RFX axisymmetric control system. In
the outer loop the position regulator produces a correction of the feedforward vertical magnetic field calculated
according to Shafranov’s equilibrium theory on the basis of the instantaneous value of plasma current and
other equilibrium quantities such as desired major radius, minor radius, coefficient of poloidal field
asymmetry. The reference field is then transformed into 8 Field Shaping winding current references for the
inner loop. Unlike the old system, where each current was independently regulated by the corresponding
amplifier, a decoupling current control system has now been envisaged which should assure the same dynamic
behaviour of the 8 current tracking errors. The output of this loop are 8 voltage references for the amplifiers.
The whole control system has then been implemented in the non-linear FE model of RFX along with the full
poloidal field electrical circuit in order to assess its performances and robustness. Tuning of the regulator
parameters was also performed on the basis of the simulations results.
Corresponding Author:
CAVINATO MARIO
cavinato@igi.cnr.it
Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione,Corso Stati Uniti 4, I-35127 Padova
121
- C - Plasma Engineering and Control.
P3C-C-363
REAL-TIME MEASUREMENT AND CONTROL AT JET- EXPERIMENT
CONTROL
FELTON, ROBERT,
Selected also for oral presentation
O1B-C 363
Over the past few years, the preparation of ITER-relevant plasma scenarios has been the main focus
experimental activity on tokamaks. The development of integrated, simultaneous, real-time controls of plasma
shape, current, pressure, temperature, radiation, and neutron profiles, and also impurities, ELMs and MHD are
now seen to be essential for further development of quasi-steady state conditions with feedback, or the
stabilisation of transient phenomena with event-driven actions. For this thrust, the EFDA JET Real Time
Project has developed a set of real-time plasma measurements, experiment control, and communication
facilities - currently, the largest real-time experiment control facility on a Tokamak. The Plasma Diagnostics
used for real-time experiments are Far Infra Red interferometry, polarimetry, visible, UV and X-ray
spectroscopy, LIDAR, bolometry, neutron and magnetics. Further analysis systems produce integrated results
such as temperature profiles on geometry derived from MHD equilibrium solutions. The signal processing
algorithms were validated on many recorded pulses, and the systems have real-time data checks. The Actuators
include toroidal, poloidal and divertor magnets, gas and pellet fuelling, neutral beam injection and RF and
microwave beam injection. The Heating/Fuelling Operators can either define a power or gas request waveform
or select the real-time instantaneous power/gas request from the Real Time Experiment Control (RTCC)
system. The Real Time Experiment Control system provides both a high-level, control-programming
environment and interlocks with the actuators. It is capable of single-input, single-output and multiple-input,
multiple-output controls. A MATLAB facility is being developed for the development of more complex
controllers. The communications network has been critical to the successful operation of the facility. It uses
ATM (the core technology of ISDN) to multicast more than 500 signals (total), in 30 different data sets, every
few milliseconds, amongst 30 systems. The network can be extended easily and quickly, and is reliable and
robust. The EFDA Real Time project is essential groundwork for future reactors such as ITER. It involves
many staff from several institutions. The facility is now frequently used in experiments. This work has been
conducted under the European Fusion Development Agreement and is partly funded by Euratom and the UK
Engineering and Physical Sciences Research Council.
Corresponding Author:
FELTON, ROBERT
Robert.Felton@jet.uk
Euratom/UKAEA Fusion Association, Culham Science Centre Abingdon OX14 3DB UK
122
- C - Plasma Engineering and Control.
P3C-C-375
OPEN LOOP CHARACTERIZATION OF AN ACTIVE CONTROL
SYSTEM OF MHD MODES
MASIELLO ANTONIO, PER BRUNSELL(1) GIUSEPPE MARCHIORI(2) DIMITRIY YADIKIN(1)
(1)Division of Fusion Plasma Physics - Association EURATOM-VR, Alfvén Laboratory, Royal Institute of Technology, S-100
44, Stockholm Sweden (2)Consorzio RFX - Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova
Italy
In the framework of the collaboration between the research groups of RFX and T2R on the active control of
MHD modes, a digital system was installed in T2R, able to control up to 32 active coils. A state space full
electromagnetic model, already developed in view of MHD mode control in RFX, was adapted to the T2R load
assembly and saddle coils layout. It consists of a lumped parameter electromagnetic model of the saddle coil
system and a linear model of the evolution of RWMs in Reversed Field Pinch plasmas. Electrical parameters
are calculated by a FE electromagnetic model, which includes the conductive structures surrounding the
plasma column. An experimental campaign both without and with plasma has now been carried out on T2R
aimed at the electromagnetic characterization of the open loop system and to the validation of the various parts
of the model. The first series of tests was performed without plasma to evaluate the coil self-inductance and
mutual inductance with the underlying and adjacent sensors. In the presence of a passive conductor (the twolayered thin shell in T2R case), both these parameters depend on the frequency. By properly programming the
control system, tests with sinusoidal reference voltage at different frequencies were performed. In order to
analyse steady-state behaviour a voltage reference step was also provided for a time longer than the shell time
constant for the penetration of the vertical field. A satisfactory agreement was observed between the
electromagnetic parameters calculated by the FE model and those derived by processing the experimental data.
Static and rotating perturbation of different harmonic order were then applied with constant and sinusoidally
varying reference voltages. In particular, the quality of the harmonic content as a function of the number of
active coils has been analysed. A clear knowledge of the sidebands amplitude of each mode in the real
machine is mandatory for the design of the mode control system. Finally, the third group consisted of shots
with plasma. Static perturbations of different toroidal order n with a step reference voltage were applied to
study their effect on the evolution of plasma MHD modes. The purpose was to compare theoretical and
experimental growth rates and to understand how low is the threshold to start an unstable mode. This
represents a basic information to determine how clean the harmonic content must be to allow a selective mode
control.
Corresponding Author:
MASIELLO ANTONIO
antonio.masiello@igi.cnr.it
Consorzio RFX, Corso Stati Uniti 4, 35127 Padova, Italy
123
- C - Plasma Engineering and Control.
P3C-C-377
COMPARISON OF STRATEGIES AND REGULATOR DESIGN FOR
ACTIVE CONTROL OF MHD MODES
MARCHIORI GIUSEPPE, PER BRUNSELL(1), MARIO CAVINATO(2), DEMETRIO GREGORATTO(2), ROBERTO
PACCAGNELLA(2), DIMITRIY YADIKIN(1)
(1)Division of Fusion Plasma Physics - Association EURATOM-VR, Alfv¨¦n Laboratory, Royal Institute of Technology, S-100
44, Stockholm Sweden (2)Consorzio RFX - Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova
Italy
The active control of MHD modes is essential for the steady-state operation of fusion magnetic confinement
devices. Reversed field pinch experiments are characterized by an intrinsic richness of MHD modes. The
replacement of old thick shells with much thinner ones, whose time constants are typically shorter than the
pulse length, made it possible to design active control systems of RWM's in T2R and RFX. In T2R a set of
saddle coils and probes has already been installed. An experimental campaign of mode active control is now
under way whose results will be useful for improving the models and the future activity on RFX. Two main
control strategies have been considered: intelligent shell and mode control. The former is aimed at reproducing
a virtual shell by zeroing the flux through each saddle probe, the latter at investigating the possibility of a
feedback stabilization of selected MHD modes. According to the desired scheme, the real control system can
handle the magnetic field measurements of the single probes or the harmonic components produced by on-line
FFT. Since the output variables of the control model are the magnetic field harmonics, a direct analysis of the
mode control scheme is straightforward, while an inverse DFT is needed to reconstruct the field at the sensors.
Both strategies can be then implemented in the model and in the real system. The design of intelligent shell
controller is carried out taking into account the electromagnetic parameters of a coil and its coupling with the
surrounding ones. A more complex approach is needed in the case of mode control, the design being strongly
affected by the number of available actuators. In T2R some unstable modes are not within the spectrum
directly producible by the set of coils (|n|<=8), even if they can be detected by the set of probes. Two
controllers are analysed to cope with this problem: the former aims at controlling the unstable modes not
directly reachable by using the knowledge of the sidebands produced by the actuators. The latter is designed
on the basis of the developed linear model by means of the pole placement technique. In RFX, since the
number of saddle coils allows to counteract directly all the unstable modes, the sidebands are not involved in
the design of the controller. Closed loop tests on the T2R machine, for all control strategies, are presented in
the paper highlighting their different performances in terms of mode stabilization and modification of plasma
spectrum.
Corresponding Author:
MARCHIORI GIUSEPPE
giuseppe.marchiori@igi.cnr.it
Consorzio RFX, Corso Stati Uniti 4, 35127 Padova, Italy
124
- C - Plasma Engineering and Control.
P3C-C-383
ADOPTING MODERN NONLINEAR CONTROL TECHNIQUES FOR
THE PLASMA STABILIZATION ON THE NOVEL LINUX-BASED
FEEDBACK CONTROLLER OF FTU
ZACCARIAN LUCA, CRISTINA CENTIOLI (1) FRANCO IANNONE (1) MAURIZIO PANELLA (1) LUIGI
PANGIONE (2) SALVATORE PODDA (1) VINCENZO VITALE (1)
(1) Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy (2)
Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Via del Politecnico 1 - 00133 Roma,
Italy
In this paper we will report on the experimental results arising from the implementation of modern control
techniques for the optimization of the RF power coupling vs the plasma conditions in the FTU experimental
facility. These experiments are carried out by employing the open-source Linux-RTAI control system
currently running on the FTU digital feedback loop. The RF power source under consideration is a Lower
Hybrid System (LH) based on 6 gyrotrons with a nominal power output capability of 1.1 MW each. The
optimization of the coupling level between the plasma and the emitting antenna reduces the reflected power
thus maximizing the heating effects in addition to avoiding danger to the emitter (equivalently, annoying
safety shutdowns of the system). To this aim, the plasma displacement is modified by suitably adjusting the
reference input to the stabilizing feedback, according to “extremum seeking” techniques based on a modified
gradient algorithm that recently appeared in the control literature. It will be shown in the paper how this
algorithm achieves a satisfactory level of robustness with respect to measurement errors and well performs in
experimental tests, thus leading to an improved effectiveness of the RF heating system.
Corresponding Author:
ZACCARIAN LUCA
zack@disp.uniroma2.it
Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Via del Politecnico 1 00133 Roma, Italy
125
- C - Plasma Engineering and Control.
P3C-C-403
VERTICAL STABILITY OF ITER PLASMAS WITH 3D PASSIVE
STRUCTURES AND A DOUBLE LOOP CONTROL SYSTEM
PORTONE ALFREDO(1), R. ALBANESE(1), R. FRESA(2), M. MATTEI(1), G. RUBINACCI(3), F. VILLONE(3)
(1) Assoc. Euratom-ENEA-CREATE, Univ. Mediterranea, Feo di Vito, I-89060 RC, Italy (2) DIFA, Univ. della Basilicata,
Contrada Macchia Romana, I-85100 PZ, Italy (3)Assoc. Euratom-ENEA-CREATE, Univ. Cassino, Via Di Biasio 43, I-03043
Cassino (FR), Italy
The main focus of this study is the analysis of the stabilizing effect on the plasma vertical motion of the 3D
eddy currents induced in the ITER Vacuum Vessel and Outer Triangular Support and the comparison with the
predictions obtained by means of simplified 2D models. This study includes – firstly - the derivation of linear
models describing the dynamics of the n=0 plasma displacements around the main ITER equilibrium
configurations. These models are derived by linearising the Grad-Shafranov equation about few key
equilibrium configurations, thus granting the consistency of the derived linear models with the MHD
equilibrium constraint. Secondly, an assessment of the effects of the 3D geometry of the Vacuum Vessel,
Blanket and Outer Triangular Support is carried out, with particular emphasis on the effects of the upper and
lower ports on plasma stability margin, growth time, minimum stabilization voltage and control loop phase
margin. At last, the performances of the present ITER control system (single loop) are assessed and compared
to those of an upgraded system (double-loop) that is here proposed to improve the stability domain of the
ITER plasmas forecasted.
Corresponding Author:
PORTONE ALFREDO(1)
alfredo.portone@tech.efda.org
EFDA-CSU, Boltzmannstrasse 2, D-85748 Garching, Germany
126
- C - Plasma Engineering and Control.
P3C-C-408
THE BASIC METHODS FOR UNDERSTANDING OF PLASMA
EQUILIBRIUM TOWARD ADVANCED CONTROL
KURIHARA KENICHI, KAWAMATA YOUICHI, SUEOKA MICHIHARU, HOSOYAMA HIROMI,
YONEKAWA IZURU, SUZUKI TAKAHIRO, OIKAWA TOSHIHIRO, IDE SHUNSUKE, JT-60 TEAM
Japan Atomic Energy Research Institute Naka Fusion Research Establishment 801-1 Mukoyama Naka-machi Naka-gun
Ibaraki-ken, 311-01, JAPAN
Since tokamak magnetic fusion research has just made a step forward to an international collaborative project
ITER, the existing tokamaks including JT-60 are expected to explore more advanced operation scenarios. To
test those scenarios in the JT-60 experiment, the basic methods for understanding of plasma equilibrium
applicable to the ITER have been developed. Some of them have been accomplished, and the other are being
conducted as follows: (1) A complete plasma shape is precisely reproduced in real time. Plasma entire current
and the weight center (near magnetic axis) of the current are provided also in real time. (2) Eddy current
effects are considered for shape reproduction. (3) A plasma current profile in the poloidal cross-section is
reproduced in real time through a new advanced algorithm. (4) For long-pulse DT operation in ITER, pick-up
coil sensors are equipped near a plasma, while absolute magnetic field sensors are placed in a distance from a
plasma to avoid high neutron irradiation. A method is developed to correct the drifted signal of the integrator
for a pick-up coil by employing distant sensor signals. In the symposium, those methods will be explained in
detail with the experimental results at JT-60, and the calculation in the ITER geometry, including the
remaining problems. On the basis of such discussion, we would like to envisage a future of plasma equilibrium
control toward ITER and a fusion power plant.
Corresponding Author:
KURIHARA KENICHI
kurihark@naka.jaeri.go.jp
Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, 801-1 Mukoyama Naka-machi,
Naka-gun, Ibaraki-ken, 311-0193 Japan
127
- C - Plasma Engineering and Control.
P3C-C-457
XSC PLASMA CONTROL: TOOL DEVELOPMENT FOR THE SESSION
LEADER
AMBROSINO GIUSEPPE, R. ALBANESE(2) M. ARIOLA A. CENEDESE(3) F. CRISANTI(4) G. DE TOMMASI M.
MATTEI(2) F. PICCOLO(5) A. PIRONTI F. SARTORI(5) F. VILLONE(6) AND JET-EFDA CONTRIBUTORS
(2)Assoc. Euratom-ENEA-CREATE, Univ. Mediterranea Reggio Calabria, IT (3)Assoc. Euratom-ENEA-Consorzio RFX, IT
(4)Assoc. Euratom-ENEA-Frascati, IT (5)EURATOM-UKAEA Fusion Association, UK (6)Assoc. Euratom-ENEA-CREATE,
Univ. Cassino, IT
A JET Enhancement was aimed at designing and implementing a new model-based shape controller (XSC, i.e.,
eXtreme Shape Controller) able to operate with high elongation and triangularity plasmas. In 2003 the XSC
has been implemented on a new hardware architecture and successfully tested in a various experiments. The
use of XSC implies a number of steps, which at present are not automated and therefore imply the involvement
of several experts. The first step is plasma modelling, based on linear (CREATE-L) and nonlinear (CREATENL) tools, which call for: - definition of target plasma configuration in terms of plasma current, shape and
main plasma current profile parameters (poloidal beta and internal inductance); - determination of a set of
poloidal field currents; - production of a linearized model. The second step is the controller design, based on
the CREATE-XSC-GEN tool, which uses SVD (singular value decomposition) to find the best combination of
currents to obtain specific changes in the shape, requiring: - selection of weight for plasma shape parameters
(plasma-wall gaps) and actuators (circuit currents); - selection of SVD tolerance; - determination of controller
gains; - test of controller robustness and definition of operational space. The third step is the creation of the
configuration file. The CREATE-EGENE tool creates a set of similar configurations each with a related set of
equilibrium currents using as input both the linear model and the controller design and produces the
configuration file, containing the information needed for the Level 1 XSC session leader interface. The fourth
step is the use of Level 1 XSC session leader interface, offering the possibility of picking up a baseline shape
and adjusting it using visual tools, showing the nominal currents needed for each value of poloidal beta and
internal inductance.
Corresponding Author:
AMBROSINO GIUSEPPE
ambrosin@unina.it
Associazione Euratom-ENEA-CREATE, Dip. di Informatica e Sistemistica, Univ. Napoli Federico II, Via
Claudio 21, I-80125 Napoli, Italy
128
- C - Plasma Engineering and Control.
P3C-C-463
A FLEXIBLE AND REUSABLE SOFTWARE FOR REAL-TIME
CONTROL APPLICATIONS AT JET
GIANMARIA DE TOMMASI, FABIO PICCOLO(1) FILIPPO SARTORI(1) JET-EFDA CONTRIBUTORS(2)
(1)EURATOM-UKAEA Fusion Ass. Culham Science Centre, Abingdon, OX143EA, UK (2)Work performed under EFDA and
partly funded by EURATOM and the UK Engineering and Physical Sciences Research Council.
At the heart of the JET machine, several real-time control, measurement and protection systems collaborate in
order to satisfy the most sophisticated experimental needs. A group of these systems shares the same software
architecture: the plasma shaping system “Extreme Shape Controller”(XSC), parts of the “Vertical
Stabilisation” control and acquisition, the “Error Field Coil Controller”. Also many can be found among the
“Real Time Data Network” (RTDN) processing nodes: the one measuring plasma internal parameters, the
density and q profile estimator, two different real-time equilibrium codes, a neural network based disruption
prediction code and a Matlab script real-time executor. The fast growth of the JET real-time control network
and the resulting need of shorter development cycle were the triggers that started the development of the
“JETRT” software. Initially just designed to help producing RTDN control node PCs, because of the XSC
project it finally developed in an all-purpose multi-platform real-time environment. This new architecture is
designed for maximum reuse. On average two third of the software is the same in all applications
independently from the platform. The varying part is the project specific algorithm, which is also compiled
into a separate software component, in order to achieve a separation from the plant interface code. This scheme
maximises reliability while at the same time reducing development costs. In addition it helps collaboration
with external development groups, since it allows focusing only on the novel and interesting parts of the
project without the need of knowledge about JET specific details. Finally it enables non-specialist
programmers to contribute to the real-time project. A very important feature of the architecture is that it
provides an integrated set of debugging and testing tools. The compiled task specific component can be
validated on “Matlab” by comparing it to its prototype. It can also be tested by simulating any old JET pulse,
either in a standalone fashion or real-time and integrated in a mock-up real-time network. Because of these
features it was possible to debug “XSC” on a PC Windows platform where better tools were available. Any
bug encountered while testing on the machine were successfully repeated in the laboratory in a controlled
simulation. This was the key to reducing the commissioning of Shape Controller to a few dedicated pulses
compared to the many days needed in 1994.
Corresponding Author:
GIANMARIA DE TOMMASI
detommas@unina.it
Associaz. EURATOM/ENEA/CREATE, Università di Napoli Federico II, Napoli, Italy
129
- C - Plasma Engineering and Control.
P3C-C-508
COMMISSIONING TESTS FOR CONTROL PROCESSES IN ASDEX
UPGRADE´S NEW CONTROL AND DATA ACQUISITION SYSTEM
THOMAS ZEHETBAUER, D. ZASCHE(1) T. VIJVERBERG(2) R. COLE(2) K. LÜDDECKE(2) G. NEU(1) G.
RAUPP(1) W. TREUTTERER(1)
(1)Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, D-85748 Garching (2)Unlimited
Computer Systems, Seeshaupterstrasse 15, D-82393 Iffeldorf, Germany
ASDEX Upgrade is being equipped with a new CODAC system. About 20 real-time control applications
process about 250 input signals and another 250 reference signals to perform a large number of functions for
actuator feedforward and plasma feedback control, for monitoring of machine and plasma states, and for alarm
handling, IO and reference value generation. If a control system is commissioned together with a new machine
then functions can be added consecutively and be tested iteratively while the machine´s operation range is
explored. In our case the control system must be replaced on a running machine. The complete set of functions
must be provided instantaneously at full machine parameters. To ensure that algorithmic function and real-time
performance of all application processes can be tested sufficiently prior to operation we implemented an in-situ
simulation method. The new CODAC system permits to download complete sets of trajectories of simulation
data into the processes for the measurement input and the reference value generation. Then an open-loop
simulation discharge can be started where the simulation data drives the control application processes. The
real-time process output and performance can be protocolled and analysed. Simulation data can be generated
from discharges executed and protcolled by the old control system, be computed by other sources of
simulation data, or be edited manually, to tune the simulation run to specific states of interest. In future, the
method can be enhanced for closed-loop operation with model-driven processes computing the response of
specific plasma quantities or machine components. Currently the system is used to verify application process
performance upon implementation or whenever modified, and to test control system reaction to specific failure
states.
Corresponding Author:
THOMAS ZEHETBAUER
Zehetbauer@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, D-85748 Garching
130
- C - Plasma Engineering and Control.
P3C-C-510
OPTIMIZATION OF THE IGNITOR OPERATING SCENARIO AT 11 MA
RAMOGIDA GIUSEPPE, CUCCHIARO ANTONIO (1), GALASSO GIUSEPPE (2), PIZZUTO ALDO (1), RITA
CAMILLO (1), ROCCELLA MASSIMO (1), PROF. COPPI BRUNO (3)
(1) Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy (2) Ansaldo Ricerche, Corso Perrone
25, 16152 Genova (GE), Italy (3) MIT, 02139 Cambridge (Ma), USA
Several endeavours have been made in order to minimize technological troubles in the IGNITOR design,
reducing electromagnetic loads, power supply requests and use of not very known materials. The present
design alterations comprise both the operating scenario for the poloidal field (PF) currents and the PF coils
geometry. The EM loads and the temperature gradient into the PF coils have been reduced, changing the
current distribution between the coils and introducing a more efficient grading in the most solicited ones,
avoiding the use of Dispersion Strengthened Copper foreseen in the previous design. The new reference
operational scenario for IGNITOR with a maximum plasma current of 11 MA has been modified according to
the engineering and power supply constraints. The edge safety factor was kept above 3.5 during the whole
plasma evolution, reducing the risk of low-q disruption events. This analysis and a relevant simulation of a
typical fast vertical disruption, in the most dangerous plasma conditions in IGNITOR, have been simulated
with the MAXFEA code. The obtained values are within the engineering constraints during the whole
operating scenario and the typical plasma vertical disruption. This scenario provides poloidal flux capability
enough to balance the plasma flux requirement even without relying on the effect of the bootstrap current. In
order to reduce the EM loads on plasma chamber and the related stresses due to plasma disruptions, it has been
investigated a new approach based on the possibility of mitigating these loads by using copper toroidal layers
added to the plasma chamber in proper positions. It has been found that this expedient could be quite effective
not only in increasing the time constant of the plasma displacement but even in reducing the vertical force and
its combined effect with hoop force on the vessel. This study has been carried on using MAXFEA simulations
with different location and extension of the layers, and has shown that the reduction of the vertical force on the
vessel, due mainly to the reduction of the halo current component of this force, is maximum when the copper
layers are located on the higher and lower end of the plasma chamber. In this configuration the copper layers
result to be very effective in increasing the stabilizing component of the eddy current (due to the plasma
displacement) without increasing the destabilizing one (due to the plasma current quench) and then in reducing
the forecast halo current.
Corresponding Author:
RAMOGIDA GIUSEPPE
giuseppe.ramogida@frascati.enea.it
Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy
131
- C - Plasma Engineering and Control.
P3C-C-517
PLASMA FEEDBACK CONTROLLER REORGANISATION FOR ASDEX
UPGRADE'S NEW DISCHARGE CONTROL AND DATA ACQUISITION
SYSTEM
W. TREUTTERER, T. ZEHETBAUER, V. MERTENS, G.NEU, G. RAUPP, D. ZASCHE
Accompanying the introduction of the a new discharge control and data acquisition system ASDEX Upgrade's
feedback control scheme is reworked. Since all the algorithms must be ported from the previous programming
language OCCAM into C++ and must be embedded into a completely new infrastructure environment, a good
chance to restructure also the control architecture emerged. It is not only that the diverse, historically grown
algorithmic implementations are unified and equipped with a common interface. Even more important is the
trend of clustering groups of up-to-now independent single-input-single-output (SISO) controllers into a few
multiple-input-multiple-output (MIMO) controllers. This is necessary to account better for the dependencies
and cross-couplings in the plasma process to be controlled and it provides a means for future optimisation once
the dependencies are better understood. Also the previous discharge control system had a primitive method to
adress this aspect: the control recipies. A recipe contained a valid combination of active SISO feedback
controllers, deactivating any competing controllers using the same actuators or tracking the same or dependent
control variables with different means. The new approach tries to identify such competing controllers and then
group them into MIMO controllers based on common actuators.A control mode for a whole group, switchable
in real-time, is the successor of the recipe.With this powerful tool it is not only possible to dynamically choose
the active control variables but also to change control algoritm and control parameters. In addition
configurable options for actuator specific behaviour like output limitation are provided. Additional flexibility
is offered with pre- and postprocessors plugged into the real-time signal exchange architecture in which the
controllers are embedded. These may filter measurements or pulse-modulate controller output commands e.g.
for NBI heating. Commissioning of the new control system kernel has already started. The restructured
feedback control scheme will be implemented during the second part of the commissioning phase. After the
summer maintenance break ASDEX Upgrade's new control system will be ready for operation.
Corresponding Author:
W. TREUTTERER
Wolfgang.Treutterer@ipp.mpg.de
Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching, Germany
132
- D - DIAGNOSTICS, DATA ACQUISITION AND REMOTE
PARTICIPATION.
P2C-D-22
NEUTRON ANALYSIS OF H-ALPHA AND CXRS DIAGNOSTICS OF
ITER
S.V. SHELUDIAKOV, G.E. SHATALOV, K.YU. VUKOLOV
123182, Moscow, Russia, Kurchatov sq. 1
The neutron analysis was carried out for two optical diagnostic systems of ITER, developed in RF: H-alpha
spectroscopy and Charge Exchange Recombination Spectroscopy (CXRS). Both systems are similar from the
neutron analysis point of view– they include extended cavities, along which the neutrons can penetrate up to
cryostat. H-alpha diagnostic location in the upper port of ITER sector 10 was chosen as a base option for the
detailed calculations, because its channel is most dangerous from point of view of neutron flux on window and
optical fiber waveguide. On its basis comparative analysis was done for channels of diagnostics in sectors 2
and 9. The obtained data is very important for material selection (mirror, window, fiber waveguide, fastening
assemblies), and also for definition of irradiation test conditions such as radiation fluxes and spectra. For
example, the neutron flux on the primary quartz window does not exceed 4x108 n/cm2/s. Results of irradiation
tests of quartz glass in nuclear reactor are taking into account. It is possible to use windows (lens) from KU-1
quartz glass (practically without change of light transmission in visible range) during all D-T ITER phase.
Moreover, there will be possible also to apply fiber waveguide from quartz glass if neutron flux will not be
significantly enlarged by background radiation. Neutron flux on cryostat makes 7x107 n/cm2/s even on axis of
the diagnostic channel together with background (neutron flux without any diagnostic at port), which is
permissible for requirement to access the personnel for 10 days after reactor shutdown. Anyway, the increase
of neutron flux carries local character, and the neutron flux promptly decreases from axis of the channel. Effect
of several systems integration in one port is difficult for estimating, but the carried out investigation shows the
way for construction improvement. It will allow to pick up acceptable design satisfying to the safety
requirements. For example, the introduction of the diagnostic channel additional bend allows essentially to
decrease the dose rate on cryostat. Besides a port shield plug with H-alpha diagnostic can be enlarged if
necessary up to port extension.It allows to decrease neutron flux on window two times.
Corresponding Author:
S.V. SHELUDIAKOV
geshat@nfi.kiae.ru
123182, Moscow, Russia, Kurchatov sq. 1
133
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-42
NEW CONSTRAINTS FOR PLASMA DIAGNOSTICS DEVELOPMENT
DUE TO THE HARSH ENVIRONMENT OF MJ CLASS LASERS
BOURGADE JEAN-LUC, V. ALLOUCHE, J. BAGGIO, C. BAYER, J. BEULLIER, F. BONNEAU, C. CHOLLET,
S. DARBON, L. DISDIER,D. GONTIER, M. HOURY, H.P. JACQUET, J.P. JADAUD, J.P. LE BRETON, J.L.
LERAY, P. LECLERC, I. MASCLET-GOBIN, J.P. NEGRE, J. RAIMBOURG, J. VIERNE
CEA/DIF, BP n 12, 91680 Bruyères le Châtel, France
Diagnostics developed for laser-produced plasmas have usually been designed without taking into account the
direct effects of radiated or nuclear energies emitted by the plasma itself. The future MJ class lasers (LMJ in
France and NIF in USA), now under construction, are designed to demonstrate ignition and gain of an ICF
target. A 10x gain target will emit up to 1019 DT fusion neutrons. At these neutron production levels, and
significantly below them (e.g. for pre-ignition studies), the designs of plasma diagnostics for will be
dramatically affected by these new environmental effects. These facilities will be able to focus up to 1.8 MJ
UV laser light in few nanoseconds into a mass of a less than 1/3 g. This very large power density will drive a
large amount of x-rays, debris, shrapnel. Unfortunately this colossal energy increase with respect to the
previous laser facilities (Phébus, NOVA or Omega that have or produce at maximum 30 kJ of UV laser light)
cannot be compensated for increasing the size of the emitting region, which remains nearly unchanged. The
spatial resolution of primary diagnostics must be preserved and some of the diagnostic parts will need to be
nearly as close to the plasma as in the past. The harsh environmental background induced on the diagnostic’s
active components must be taken into account not only for successful gain conditions but for many other
experiments where neutron yields are within 1014 up to 1016 n/4??? The solution to achieving the goal of
diagnostic survivability is not obvious and many new studies must be conducted to verify the diagnostic
performance under these environmental conditions. Simulations and experiments conducted at CEA/DAM
related to the preliminary studies of this harsh environment effects on diagnostic components will be
presented. Recent experiments performed at the OMEGA laser facility at the University of Rochester (USA)
within the framework of the CEA/NNSA-LLNL collaboration, including neutron effects on CCD readout,
coaxial cable, streak or framing cameras and EMP measurements, will be fruitful for the design of diagnostics
for the MJ class laser facilities diagnostic design. New measurement strategies that have been developed to
overcome these environmental difficulties will be presented. Synergies with similar environment studies
conducted for magnetic fusion diagnostic design for next ITER facility must be considered and enhanced.
Corresponding Author:
BOURGADE JEAN-LUC
jean-luc.bourgade@cea.fr
CEA/DIF - BP12 - 91680 Bruyères le Châtel - France
134
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-52
THE INTEGRATED VISUALISATION SOFTWARE FOR THE ITER IN
VESSEL VIEWING SYSTEM (IVVS)
RIVA MARCO, CARLO NERI(1), FABIO BONACCORSO(2), FEDERICO MASSAIOLI(2)
(1) Associazione EURATOM-ENEA sulla Fusione, Centro Ricerche Frascati, C.P. 65, 00044, Frascati, via E. Fermi 45 (2)
Consorzio CASPUR, Roma, Italy
The Amplitude-modulated In vessel viewing system (IVVS), developed by ENEA for the ITER machine, is a
versatile apparatus capable to scan the internal of a vessel and acquire amplitude and range data of a quasispherical view. By coupling the two images we obtain a true 3D image of the scene, with a sub-millimetric
range precision for each point scanned in the range 2-10 meters. In this article we present the integrated
software system we have developed in Frascati. This system allows to drive the image acquisition, show the
forming image during the acquisition, generate the intermediate structures for the visualisation and render the
acquired image in a 2D and/or 3D framework, allowing pan/tilt/zoom on any image portion, applying a series
of graphical filters for image enhancement, colour-coded visualisation, saving image portions in standard
formats and eventually compare portions of image with a sample data to highlight differences. Moreover the
system allows comparing a reference acquired image with any other acquired image. This would be useful
studying changes between the 3D image taken as reference respect to the current acquisition, allowing an
eventual maintenance task to be carried out.
Corresponding Author:
RIVA MARCO
riva@frascati.enea.it
Associazione EURATOM-ENEA sulla Fusione, Centro Ricerche Frascati, C.P. 65, 00044, Frascati, via E. Fermi
45
135
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-60
USING REMOTE PARTICIPATION TOOLS TO IMPROVE
COLLABORATIONS
BALME STÉPHANE, JOHN HOW, PHILIPPE LEBOURG, JEAN-MARIE THEIS, NADINE UTZEL
CEA Cadarache DRFC/STEP 13108 Saint Paul Lez Durance Cedex FRANCE
Poster presentation stephane.balme@cea.fr The Tore Supra project, by its very nature, relies on worldwide
collaboration for its physics programme. In the past this has been carried out by site visits for experiments,
backed up by use of email and telephone. However today, in the “Département de Recherche sur la Fusion
Contrôlée” (DRFC), we make substantial use of modern remote participation (RP) tools such that our distant
collaborators have a “virtual presence” in our laboratory. This includes audio/video conferencing
technologies, that can be used to participate remotely in experiments, applications code development and
meetings (H.323 standards, VRVS and multicast protocols, conference-telephones, text messaging, VNC
screen sharing coupled with interactive board, etc.), secure remote access to local computers (Citrix solution,
SSH, VPN), including code sharing (CVS), remote Matlab access to Tore Supra data via MDS+, remote
control of diagnostic instrumentation by experimentalists, broadcast of experiment states and machine
parameters screens (TSTV channels), remote monitoring with synchronisation to TS shot events via MDS+
Last year we equipped a meeting room to enable multi-institutional teams of experts to efficiently collaborate
in the fusion development effort. This room concentrates many communication and collaboration technologies
using low cost but reliable products and open source software. This effort is constantly evolving, following the
real needs of our users and has given us valuable information to maximise participation, interaction and
collaboration between the researchers for the future generation of Tokamaks. The objective of this current
report is to clearly present our installations and experience for working Tore Supra collaborations such that
interested persons can learn and profit for their own collaborations. We shall therefore emphasise the practical
applications of remote participation, rather than the principles, which are now well known. We expect a very
strong participation for Cadarache with ITER and in this perspective we wish to be able to share our practical
experience in RP techniques.
Corresponding Author:
BALME STÉPHANE
stephane.balme@cea.fr
CEA Cadarache DRFC/STEP 13108 Saint Paul Lez Durance Cedex FRANCE
136
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-74
CALORIMETRY MEASUREMENTS DURING HIGH ENERGY
DISCHARGES AT TORE SUPRA
CHANTANT MICHEL, B.BEAUMONT, P.BIBET, A.EKEDAHL, A.MARTINEZ, JC.VALLET
Association Euratom-CEA DRFC CEA/Cadarache F 13108 St Paul Lez Durance
One particularity of Tore Supra is that the Plasma Facing Components (PFCs) are actively cooled, allowing
research on long duration and high energy discharges. During a discharge, a large part of the energy is
dissipated in the heating generators (ICRH, ECRH and LHH), auxiliaries and transferred to the primary
cooling loops (B50 and B60). The energy which is injected in the plasma is totally recovered by the PFCs and
transferred to an other primary loop (B30). The primary loops are cooled via heat exchangers by a secondary
loop. The previously existing limitations of the complete Tore Supra cooling system were due to the heat
exchange performance of the secondary loop cooling towers. Schematically, during a long duration and high
energy discharge, only the PFCs primary loop is cooled by the secondary loop, while the hot water from the
heating generators primary loops is stored in tanks which were inserted at the outlet leg of the generators. At
the end of the discharge, the hot water stored in the tank is cooled when the PFCs primary loop has recovered
its initial temperature. In early 2003, the modifications mentioned above were completed and the loops were
fully operational. During the 2003 experimental campaign significant results were obtained, particularly during
non inductive discharges with an injected energy by the Lower Hybrid Current Drive (LHCD) system of up to
1.1 GJ. During the discharge, the circulation of one part of the B30 flow is monitored in a heat exchanger to
achieve a good stability of the water temperature at the tokamak inlet. This condition is particularly important
to measure accurately the energy balance of the tokamak by calorimetry. For the long and high energy
discharges, the agreement between the energy measured by the calorimetry on the PFCs cooling loop and the
HF energy is very good (90 to 95%). The energy radiated by the plasma can be measured from the calorimetry
of the panels protecting the vacuum vessel. Furthermore, calorimetry measurements on the B60 cooling loop
allow to assess the global LHCD system efficiency (20%), which was not requested to operate at its maximum
efficiency in these experiments. The paper will present the main characteristics and operation modes of the
Tore Supra cooling loops. It will also give the description of the calorimetry diagnostic and a detailed analysis
of the PFCs calorimetry measurements for discharges with injected energies in the range 400 MJ to 1.1GJ.
Corresponding Author:
CHANTANT MICHEL
chantant@drfc.cad.cea.fr
Association Euratom-CEA DRFC CEA/Cadarache F13108 St Paul Lez Durance
137
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-80
REAL-TIME MULTIPLE NETWORKED VIEWER CAPABILITY OF
THE DIII-D EC DATA ACQUISITION SYSTEM*
PONCE, D., Y.A. GORELOV
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
A data acquisition system (DAS) which permits real-time viewing by multiple locally networked operators is
being implemented for the electron cyclotron (EC) heating and current drive system at DIII-D. The DAS is
expected to demonstrate performance equivalent to standalone oscilloscopes. Participation by remote viewers,
including throughout the greater DIII-D facility, can also be incorporated. The EC system at DIII-D consists of
five 1 MW class gyrotrons. Operational performance is monitored by collecting beam voltage and current,
generated rf, and tube pressure waveforms, among other signals. These are currently disseminated in three
ways. First, operators can view a subset of the available signals on traditional oscilloscopes. Second, legacy
CAMAC digitizers are used to capture the signals and download them to the DIII-D DAS between DIII-D
shots. Third, operators can view signals as they are acquired using compact PXI based digitizers. This last set
of digitizers is used in the real-time system. The real-time system uses 1 computer controlled DAS per
gyrotron. Each computer acquires 8 channels simultaneously sampled at up to 70 kHz per channel.
Additionally, the computer acquires and buffers 4 channels of fault signals simultaneously sampled at 15 MHz
and 16 channels of coolant temperature and flow used for calorimetry. All signals can be viewed on the
acquiring computer. The 8 real-time channels per system will be distributed to multiple remote viewers. Each
DAS computer sends its data to a central data server using individual and dedicated 100 Mbps fully duplexed
Ethernet connections. The server has a dedicated 10,000 rpm hard drive for each gyrotron DAS. Selected
channels can then be reprocessed and distributed to viewers over a standard local area network (LAN). They
can also be bridged from the LAN to the internet. Calculations indicate that the hardware will support realtime writing of each channel at full resolution to the server hard drives. The data will be re-sampled for
distribution to multiple viewers over the LAN in real-time. Hard drives will hold about a weeks worth of data.
Archives will be kept on DIII-D servers and locally on a terabyte DVD/CD changer. The hardware for this
system is in place. The software is under development. This paper will present the design details and up-todate performance metrics of the system. *Work was supported by the U.S. Department of Energy under DEFC02-04ER54698.
Corresponding Author:
PONCE, D.
ponce@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
138
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-91
EXPERIMENTAL STUDY OF RADIATION-INDUCED CURRENTS IN
COPPER AND STAINLESS STEEL CORE MINERAL-INSULATED
CABLES IN THE BR2 RESEARCH REACTOR
VERMEEREN LUDO,
To monitor the shape and position of the plasma boundary in ITER, in-vessel magnetic coils made of mineral
insulated (MI) cables will play a crucial role. In view of the high radiation levels in ITER and the projected
long pulse duration, radiation-induced electromotive force (RIEMF) effects (possibly in combination with
thermo-electric effects) might jeopardize a proper functioning of magnetic diagnostic devices: currents will be
generated between the core wires and the sheaths of the MI cables and spurious voltage differences can arise
along the core wires. An experimental study of the RIEMF effect on MI cables has been performed in the
framework of the development of magnetic diagnostics for ITER. In this study, six 1 mm diameter MgO
insulated cables (with copper and stainless steel cores) were exposed to the combined neutron-gamma field of
the BR2 reactor (thermal neutron fluxes up to 3E14 n/(cm²s), fast neutron fluxes up to 2E14 n/(cm²s) and
gamma heating rates up to 3.5 W/g). The currents between the inner conductors and the sheaths were recorded
as a function of the position of the irradiation rig and compared to theoretical predictions based on a Monte
Carlo code. Promptly induced currents and delayed currents due to radioactive decay of various activation
products were analyzed separately. Special attention was paid to the dependence of the induced currents upon
the environment and even upon the relative position of the cables and the surrounding materials with respect to
the radiation source. A systematic trend in this dependence was observed. In accordance with the theoretical
predictions, the copper-core cables show significantly higher induced currents than the stainless steel-core
cables due to the beta decay of 66Cu formed after neutron capture by 65Cu. Total equilibrium currents for
straight 1 mm diameter cables passing once through the 76 cm high BR2 core ranged from 0.06 to 0.1 µA for
copper-core cables and from 0 to 0.02 µA for stainless steel-core cables. Future work in this field will
concentrate on the RIEMF voltages developing along the MI cables and on the influence of thermal gradients
on RIEMF effects.
Corresponding Author:
VERMEEREN LUDO
lvermeer@sckcen.be
SCK-CEN, Boeretang 200, B-2400 Mol, Belgium
139
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-98
TORE-SUPRA INFRARED THERMOGRAPHY SYSTEM, A REAL
STEADY STATE DIAGNOSTIC.
GUILHEM DOMINIQUE, B.BERTRAND, C.DESGRANGES, M.LIPA, P.MESSINA, M.MISSIRLIAN,
C.PORTAFAIX, R.REICHLE, H.ROCHE, A.SAILLE
Association Euratom-CEA, CEA/DSM/DRFC CEA-Cadarache 13108 Saint Paul Lez Durance (France)
Within the framework of the CIEL and CIMES projects, a flat (7m²) toroidal pumped limiter (TPL) and 5
radio-frequency (RF) antennae, all actively cooled, have been implemented in Tore Supra tokamak. The TPL
is designed to extract up 15 MW of conducted power with a maximum power density of 10MW/m². From our
past experience with such an entirely actively cooled machine [1] we can conclude that the main limitations
may come from 1) defects present on elements due to pre-existing cracks or bonding defects, 2) suprathermal
particles impacts such as fast ions/electrons trapped in magnetic field ripple, 3) unexpected events like
“superbrillance” [2]. Such events may also happen in ITER [3]. We have developed an infrared (4.4 to 4.6 µm)
thermographic system (20ms time resolution, 9mm spatial resolution). It is made of a set of 7 endoscope
bodies, each equipped with 3 viewing lines : 2 x 30 sectors of the TPL (merged outside of the port and viewed
by one IR camera) and 1 viewing line for 1 antenna (viewed by another camera). Each optical line is made of
28 lenses (Ge, ZnS, ZnSe), 4 mirrors and 2 sapphire windows. The lenses are embedded in an actively cooled
(25 C) stainless steel tube. The outer jacket of the endoscope body is at the same controlled temperature as the
tokamak (~150 C) to prevent it from becoming a cold point under vacuum. This actively cooled jacket is
designed to withstand disruptions as well as radiated power deposition and conducted power from electron
ripple losses, thanks to an actively cooled CuCrZr protection plate. The sapphire window closest to the plasma
is joined via a OFHC copper ferrule to the actively cooled CuCrZr endoscope head, able to withstand in steady
state, the radiated plasma power loss (up to 0,2MW/m²) [4]. These sapphire windows (3 per endoscope body)
are protected during the conditioning of the machine (glow discharges or boronisation) by a rotating radiative
CFC shutter (860 C max for 10MW of radiated plasma power loss). The inner actively cooled tube supporting
the lenses can be removed without breaking the inner vessel vacuum. The successful operation since 2002 has
shown that the dimensioning had been correctly done. [1] Equipe TORE-SUPRA. Fusion Technology, vol 29,
pp. 417-448, 1996 [2] D.Guilhem, A.Seigneur ; JNM 196-198 (1992) 759-764 [3] F.C.Schuller and
A.A.Oomens; Fusion Engineering and Design 22 (1993) 35-55 [4] M. Lipa and al , Fusion engineering and
design 61-62 (2002) 801-806.
Corresponding Author:
GUILHEM DOMINIQUE
dominique.guilhem@cea.fr
Association Euratom-CEA, CEA/DSM/DRFC/SIPP/CEC, CEA-Cadarache, 13108 St PAUL lez DURANCE,
FRANCE
140
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-104
NEW INSTRUMENTS FOR ADVANCED NEUTRON EMISSION
SPECTROMETRY DIAGNOSIS OF D AND DT PLASMAS AT JET
JAN KÄLLNE, S.CONROY, G.ERICSSON, M.GATU JOHNSON, L.GIACOMELLI, W.GLASSER, G.GORINI(1),
H.HENRIKSSON, A.HJALMARSSON, M.JOHANSSON, J.KÄLLNE, S.POPOVICHEV(2), E RONCHI,
H.SJÖSTRAND, J.SOUSA(3), E.SUNDÉN ANDERSSON, M.TARDOCCHI(1), M.WEISZFLOG, JET-EFDA
CONTRIBUTORS
(1)INFM, Univ. Milano-Bicocca and Inst. di Fisica del Plasma, ENEA-CNR Assoc., Milan, Italy (2)JET, Culham Science
Centre, ABINGDON, UK, UKAEA Assoc. (3) Assoc. IST, Centro de Fusão Nuclear, Inst. Superior Técnico, 1049-001 Lisboa,
Portugal.
The magnetic proton recoil (MPR) spectrometer was developed for use on the deuterium-tritium experiments
at JET. This demonstrated new capabilities of neutron emission spectroscopy (NES) diagnosis of plasmas far
beyond expectations, especially, with regard to aspects of importance for ITER. This success gave the impetus
to including two NES projects in the enhancement program at JET, which is reported on here. One NES
project involves an upgrade of the MPR (MPRu) to facilitate measurements with higher precision and
enhanced suppression of background radiation. The latter will allow diagnostic utilization of the weakest
spectral components set by the statistics of the measurement. Similarly, the MPRu operates over the entire
fusion neutron energy range (0-20MeV), which allows diagnosing D plasmas; this was not possible with the
MPR. The MPRu uses a focal plane detector array based on the phoswich technique, where laminated
scintillators of different time response characteristics are used. The other project entails the development of a
time-of-flight spectrometer optimized for rate (TOFOR), which, statistics-wise, should match the performance
of the MPR for DT plasmas; i.e., operation in the count rate range of hundreds of kHz which is two orders of
magnitude faster than previously attained for D plasmas. This is achieved by careful design of the geometries
of the scintillation detectors making up the TOFOR instrument; the design is derived from extensive
simulations of the neutron response of the detectors and their light emission and transport characteristics. Both
spectrometers will be fully calibrated before installation with reference to certain working points. A
rudimentary control and monitoring system was used on the MPR and this has now been further developed.
This together with a state of the art communication and data acquisition system permits experiments with
MPRu and TOFOR to be operated fully electronically and monitored for stability over both short (transient)
and long time periods, allowing remote examination. Time digitizers and transient recorders based on PC cards
are being developed for the first time for NES diagnostic applications. This contribution will report on the
principles of the most advanced neutron spectrometer diagnostics built for JET and their use to test reliability,
machine interface issues and diagnostic capabilities in fusion experiments that mimics ITER as close as
todays’ generation of tokamaks permits.
Corresponding Author:
JAN KÄLLNE
jan.kallne@tsl.uu.se
Dept. Neutron Research, Uppsala University, Box 525, SE-75120 UPPSALA, SWEDEN
141
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-119
SURFACE DIAGNOSTICS WITH APPLICATION OF VIDEOSCOPE ON
THE BASIS OF CU-LASER
BUZHINSKIJ OLEG, OTROSHCHENKO VLADIMIR (1)
(1) State Scientific Center Troitsk Institute for Innovation and Thermonuclear Researches, 142190 Troitsk Moscow reg.
Russia
Research of materials surface in extremely stressed conditions under effect of plasma, welding, melting,
radiative radiation represents the great scientific interest, but has a series of formidable experimental
restrictions. Use of light amplification effect in a visible range in active optical medium allows successfully to
overcome them, in particular, in conditions of intensive plasma and radiative influence. For this goal it is
possible to apply a laser with high amplification coefficient,working practically in a single-pass mode. At the
illumination of a researched surface area the laser peak intensity should to exceed significantly a background
radiation in the given solid angle. For obtaining of the observed surface image it is necessary to collect a laser
radiation scattered by a surface area, selectively to amplify it and to focus on the receiving part of a recording
device. If laser works in a pulsed operation mode it is possible to receive in real time a spatially temporal
surface image. In the work a spatially temporal diagnostics for surface research using videoscope on the basis
of Cu-laser is presented. The typical sizes of a researched surface area essential to the given method are
determined. The optical scheme of diagnostic device is represented. The description of device block diagram is
given. The limiting spatially temporal resolution is determined. Taking into account of optical properties of
researched medium the application of chosen observation method is substantiated. The results of experimental
researches of the videoscoping system are submitted. In experiments the copper vapor laser with
characteristics was applied: -visible radiation at green (510.6 nm) and yellow (578.2 nm) wavelengths; -pulse
repetition rate - 12 kHz; -average radiation power - 4 W; -generation pulse duration - 15 ns; -peak radiation
power - 80 kW; -laser beam diameter - 12 mm; -laser tube length - 500 mm; -laser radiation divergence - 1
mrad.
Corresponding Author:
BUZHINSKIJ OLEG
buzh@triniti.ru
State Scientific Center Troitsk Institute for Innovation and Thermonuclear Researches, 142190 Troitsk Moscow.
reg. Russia
142
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-120
NEW MANAGING SYSTEM OF A LARGE AMOUNT OF IMAGES ON
TORE SUPRA
BURAVAND YVES, L. DUCOBU, D.GUILHEM, H.ROCHE, J.M.TRAVERE, N.UTZEL
CEA Cadarache DRFC-STEP BP n 1 13108 Saint-Paul Lez Durance France
Poster presentation After a major replacement of all internal components (terminated in 2002), Tore Supra has
re-started operation with a new set of actively cooled components. The safety of these components relies on
infrared (IR) thermography. Up to twelve IR cameras will be used to survey nearly all the high flux target
zones which are essentially the 360 toroidal pump limiter and the 5 power injection HF antennas. For dealing
with the large amount of data produced (~ 8Mbytes/s/camera) a two stages system has been implemented. At
the first stage, IR data are acquired and stored in separate units (Industrial PCs) that can serve two cameras.
Between shots, a few PCs, running a dedicated software (ShotPlayer), have a fast access to these data for
analyses purposes. All these workstations are linked by a dedicated gigabit network. Each night, data stored on
the acquisition units are compressed using a free JPEG library. Compression parameters are adjusted in order
to have an acceptable degree of loss of information and a rather good compression ratio (~10). These
compressed images are then moved to a dedicated IR data server but the original films are kept here for 2
weeks before being deleted, except if required by a physicist to keep the raw data. The IR server has an
extensible 700 Go mass storage space and runs two software servers, one dedicated to the ShotPlayer clients,
and another one for in-depth analysis via the Matlab® program. Physicists can access raw compressed images
through a multithreading, overlapped I/O server, decompress and convert them in absolute temperature on
their workstations using functions we have developed and integrated into the Matlab® environment.
Acquisition units, network, and the ShotPlayer workstations clients were tested during the 2001 Tore Supra
campaign with only one IR camera. Because of its modularity, the acquisition system has easily evolved
during the 2002-2003 experiments; 8 cameras will be in operation in 2004. Also, the whole long term storage
and retrieval system is now validated. Techniques and methods used to realize this imaging diagnostic give us
invaluable information that will also be valuable for defining future generation of imaging data acquisition
systems.
Corresponding Author:
BURAVAND YVES
yves.buravand@cea.fr
CEA Cadarache DRFC-STEP BP n 1 13108 Saint-Paul Lez Durance France
143
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-132
RADIATION RESISTANT BOLOMETERS USING PLATINUM ON
AL2O3 AND ALN
M. GONZALEZ, R. VILA, AND E.R. HODGSON
Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain
Present day JET type bolometers using gold meanders on a mica substrate have been considered for use in
ITER. However in-reactor tests at JMTR (Japan) demonstrated that they suffer radiation damage due to
transmutation of gold into mercury, and detachment of the meander from the substrate possibly related to this
alloy change and/or substrate swelling. The formation of a gold / mercury alloy caused a dose dependent
increase in the bolometer meander resistance, and the detachment of the meanders was accompanied by loss of
electrical continuity. As a result work was undertaken to examine alternative more radiation resistant
substrates, together with the substitution of evaporated gold by sputtered platinum. From these initial studies
already reported, commercially available sheets of alumina and AlN were selected for further testing. The
radiation resistance of these prototype bolometers with sputtered platinum meanders is now being
characterised. As a first step their behaviour as a function of temperature during electron irradiation has been
examined, before neutron irradiation tests are performed. The results for the meander resistance as a function
of thermal cycling and radiation dose are considered to be satisfactory. In parallel a detailed study is being
carried out to compare evaporated versus sputter deposited platinum on the ceramic substrates to minimise
detachment problems. An analysis on the basis of the quality of the platinum sensor adhesion and the metalsubstrate interface features is performed to optimise the bonding in an attempt to obtain bolometers of
improved radiation resistance. The role of glow discharge atmosphere, platinum source, and other relevant
process parameters will be presented and discussed. In addition to compare with the original gold on mica
bolometers, platinum on mica has been prepared and tested for comparison.
Corresponding Author:
M. GONZALEZ
MGV@CIEMAT.ES
Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain
144
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-142
TEMPERATURE DEPENDENCE OF THE TRANSMISSION LOSS IN
KU-1 AND KS-4V QUARTZ GLASSES FOR THE ITER DIAGNOSTIC
WINDOW
NISHITANI TAKEO, SUGIE TASUO, MORISHITA NORIO, YOKOO NORIKO
In the ITER diagnostics, optical windows will be installed at the end of diagnostic ports, where the ionizing
dose is expected to be less than several MGy for the ITER lifetime. Quartz glasses of KU-1 and KS-4V are
candidate window materials for the ITER optical diagnostics in UV and visible range. KU-1 and KS-4V have
been developed as radiation resistant quartz glasses in Russia. KU-1 is characterized by a high OH content. On
the other hand, KS-4V is low OH and low chlorine contents. Quartz glasses have relatively large sensitivity in
UV range for radiation. The temperature dependence of the radiation induced transmission losses in UV range
has been investigated under gamma-ray irradiation for the KU-1 and KS-4V quartz glasses up to 10 MGy. The
Co-60 irradiation facility of JAERI/Takasaki was used for this irradiation test. The KU-1 and KS-4V quartz
glass samples with the size of 16 mm in diameter and 8 mm in thickness were irradiated at room temperature,
100, 200 and 300 C. Dose rate was 0.46 Gy/s until 1.1 MGy, and 2.6 Gy/s from 1.1 MGy to 10.1 MGy. The
transmission spectra of the windows were measured with a spectrometer once a week during irradiation pause.
It was confirmed that there are no significant loss in wavelength range longer than 350 nm for both window
materials. Transmission loss in KU-1 is larger than that in KS-4V under temperature below 200 C. KU-1 has
large temperature dependence. On the other hand, temperature dependence is not clear in KS-4V above 100 C.
At 300 C, transmission loss in KU-1 is smaller than that in KS-4V. Prominent recover of the transmission loss
was not observed after irradiation for each sample. In the transmission spectrum of KU-1, two absorption
peaks were identified; one was E’-center at 215 nm and the other is from the non-bridging oxygen hole center
(NBOHC) at 260 nm, which is common feature of the quartz glass for the gamma-ray irradiation. In KS-4V,
absorption of NBOHC is not clear, but the absorption peak from oxygen deficient center was observed at 245
nm, which suggests that the defect production mechanism is different in KU-1 and KS-4V. From the
transmission loss point of view, KS-4V is better window material than KU-1 at temperature below 200 C.
KU-1 is available at 300 C.
Corresponding Author:
NISHITANI TAKEO
nisitani@naka.jaeri.go.jp
JAERI, 2-4 Shirakata-shirane,Tokai-mura, Naka-gun, Ibaraki 319-1114, JAPAN
145
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-159
THERMAL AND NEUTRON TESTS OF MULTILAYERED DIELECTRIC
MIRRORS
ILIA ORLOVSKIY, KONSTANTIN VUKOLOV
Optical diagnostics for ITER will require mirrors for transmission of radiation from plasma to detectors.
Apparently, plasma-facing (first) mirrors should be metallic because they will be exposed to significant fluxes
of fast neutrons and charge exchange atoms (CXA). At the same time, secondary mirrors are expected to take
less neutron fluxes and no CXA so possibly they can be made of material with higher reflectance. One way to
improve the transmittance is to use multilayered dielectric mirrors as the secondary mirrors. Multilayered
dielectric mirrors are made up of alternating layers of materials of different refraction and their reflectance
reaches 100% in a certain range of wavelengths. Then there is an open question of capability of such mirrors
for operating under ITER conditions which are neutron fluxes up to 10E12 n/cm2s and temperatures of 150C
to 200C. The goal of our experiment is to examine some samples of dielectric mirrors for their resistance to
noted temperatures and neutron fluences to be accumulated in ITER for several years and to draw a conclusion
of a possibility of applying of dielectric mirrors to optical diagnostics of ITER. Visible-range mirrors made
from TiO2/SiO2 and ZrO2/SiO2 layers applied onto SiO2 substrates by two manufacturers were irradiated to
fluences of 10E17 n/cm2 at about 50C in nuclear reactor. The number of layers varied from 13 to 23. Neutron
fluence corresponds to a fluence to be accumulated by the fourth mirror of H-alpha spectroscopy diagnostics
for 1 year of ITER operation. The neutron test was preceded by thermal tests in which the mirrors were heated
up to 280C. Some mirrors retained their reflectance and structure integrity under heating while others did not
that can be caused by insufficient layers adhesion to mirrors substrate. All the samples did not change
significantly their reflectance under irradiation. The results obtained confirm resistance of dielectric mirrors to
neutron irradiation to high fluence. Although the resistance of mirrors to thermal load is limited by the strength
of adhesion, it is clear that sufficient adhesion can be provided by manufacturers. In addition, in spring 2004
we are going to perform the second neutron test where the same samples will be irradiated to a neutron fluence
of 10E19 n/cm2. The results will be included in the report.
Corresponding Author:
ILIA ORLOVSKIY
orlovskiy@nfi.kiae.ru
Nuclear Fusion Institute, Russian Research Center "Kurchatov Institute", 123182 Kurchatov Sq. 1, Moscow,
Russia
146
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-166
LASER DAMAGE INVESTIGATIONS OF CU MIRRORS
ALEXEY GORSHKOV, IGOR BEL’BAS MIKHAIL MASLOV VLADIMIR SANNIKOV KONSTANTIN VUKOLOV
RRC "Kurchatov Institute", Kurchatov sq.1, 123182 Moscow, Russia
Laser tests were performed on Cu mirrors as “first mirror“ prototypes for laser diagnostics for ITER. Data on
laser damage thresholds under the influence of frequency-operated pulsed YAG laser (1064 nm wavelength,
12,5 Hz repetition rate, 10-30 mJ per pulse energy, 26 ns pulse duration) were obtained for single laser shot
and after about 1.5x10^5 laser shots. The output beam of the laser operating in the TEMoo mode has a
Gaussian profile. The experiments were carried out with using three types of the mirrors: one – diamondturned Cu mirror, second – diamond-turned substrate with Cu coating and third - reflection grating on the Cu
coating mirror. The single shot damage thresholds were measured and were equal to 27±5.5 J/cm^2 for
diamond-turned mirror, 18.7 ± 3.7 J/cm^2 for Cu coated mirror and 12.3 ± 2.5 J/cm^2 for the Cu grating. The
lifetime of the Cu mirror under multiple pulse laser irradiation was studied. Diffusion scattering was used as a
monitor of mirror surface condition. The degradation of copper mirrors under multiple pulse laser irradiation is
described with satisfactory accuracy by a predictive model for multipulse laser damage of metal mirrors up to
1.5x10^5 laser pulses. This model relates multiple-pulse damage that accumulated on metal surfaces to the
thermal stress field induced by the laser pulse. Keywords: Optical diagnostic of the plasma; Cu mirror; Laser;
Laser damage threshold; Diffusion scattering.
Corresponding Author:
ALEXEY GORSHKOV
gorshkov@nfi.kiae.ru
RRC
147
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-194
DESIGN OF LOST ALPHA PARTICLE DIAGNOSTICS FOR JET*
BAEUMEL STEFAN, A. WERNER(1), R. SEMLER(1), S. MUKHERJEE(1), D.S. DARROW(2), R. ELLIS(2), F.E.
CECIL(3), V. KIPTILY(4), L. PEDRICK(4), J. GAFERT(4) AND JET-EFDA CONTRIBUTORS(4)
(2)Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543-0451, USA (3)Colorado School of Mines, 1500
Illinois St., Golden, CO 80401, USA (4)EFDA-JET, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK
Two diagnostics for the measurement of lost alpha particles are being fabricated for the Joint European Torus
(JET). These diagnostics consist of a scintillator based probe [1,2] near the midplane and a poloidal array of
five sets of thin foil Faraday collectors, which replaces the existing Kalpha-1 diagnostic [3,4]. Both systems
are capable of measuring fast ions, fusion products and ICH tail ions. The Faraday cup array will measure the
particle loss at multiple locations at a rate of 1 kHz, while the scintillator probe will be capable of measuring
the pitch angle and energy distribution of the lost ions with a time resolution of 2 kHz. The dynamic range of
the Faraday cups, which consist of a stack of 2.5 microm thin Ni foils, will be from 1 nA/cm2 to 100
microA/cm2. The dynamic range for current measurements with the scintillator probe will be from 10pA/cm2
to 1 microA/cm2. Not all Faraday cups will have the same energy resolution. It will range from 15-25% for 3.5
MeV alpha particles. The scintillator probe is designed to detect fast ions with pitch angles up to 86 and
provides a pitch angle resolution of about 5 degrees. The gyroradius resolution of the scintillator probe will be
about 15%. The scintillator material will be P56 (Y2O3:Eu) will be used, which is luminous up to about 400
C. In order to maximize the signal to noise ratio, the detectors of both diagnostics are located as close to the
face of the poloidal limiter structure as feasible (5 mm). Therefore both diagnostics protrude quite far into the
vessel. Due to Halo- and Eddy currents that lead by the interaction with the background magnetic field to
enormous forces onto the diagnostics, considerations of structural stability were a major concern of the design.
Since the probe is located beside a neutral beam injector that deposits a significant heat load of 13 MW/m2 on
the side of the probe, it will be actively cooled. Experience in operating both diagnostics in a high temperature
and high radiation environment will be valuable in preparation for the design of similar diagnostics for future
fusion devices. 1. S.J. Zweben et al. Nucl. Fusion 30, 1551 (1990). 2. A. Werner et al. Rev. Sci. Instruments
72, 780 (2001) 3. F.E. Cecil et al. Rev. Sci. Instruments 70, 1149(1999). 4. O.N. Jarvis et al. Fusion
Technology 39, 84 (2001) *This work is supported by U.S. Department of Energy Contracts DE-AC0276CH03073 and DE-FG03-95ER54303 and conducted under EFDA by IPP, PPPL and CSM.
Corresponding Author:
BAEUMEL STEFAN
baeumel@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Wendelsteinstr. 1, 17491 Greifswald, Germany
148
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-202
DEVELOPMENT OF THE PHASE COUNTER WITH THE REAL-TIME
FRINGE JUMP CORRECTOR FOR INTERFEROMETER ON LHD
YASUHIKO. ITO, KENJI. TANAKA, TOKIHIKO. TOKUZAWA, TSUYOSHI. AKIYAMA, SHIGEKI. OKAJIMA AND
KAZUO. KAWAHATA
Dept. of Eng. & Tech. Servises National Institute for Fusion Science, Oroshi-cho, Toki, GIFU, 509-5292, Japan
In a laser heterodyne interferometer system, a phase counter is used to measure plasma electron density from
the beat signals of interferometer outputs. One of the important problems in multi-fringe phase detection is the
fringe jump error. The error is caused by decreased signal to noise ratio of the beat signal, when the electric
noise is increased or the signal amplitude is decreased due to refraction of the probe beam. The fringe jump
occurs also due to the spike-like stray pick-up when the discharge of ion source of neutral beam injector (NBI)
breaks. The fringe jump causes severe problem for the data analysis and operation of density feedback. We
developed the circuit, which called the AFJC (Automatic Fringe Jump Corrector), to compensate the fringe
jump automatically. The AFJC circuit was developed to improve fringe jump correction work. The functions
of the circuit are fringe jump detection and change incorrect fringe number of the phase counter to the correct
fringe number stored before the fringe jump. The circuit can process fringe jump correction within 3us. The
circuit is integrated on a CPLD (Complex Programmable Logic Device), which added to the conventional
phase detection circuit with the following specifications: 1MHz input beat frequency, 31 fringes phase
detection range and 10us phase resolution. The circuit was installed on the FIR laser interferometer and was
tested in the electron density measurements of LHD plasma. As a result, the fringe jumps caused by the H2
pellet injection and the radiation collapse can be corrected, the probabilities of correction were approximately
70% and 20% respectively. However, the fringe jump caused by high-density plasma measurement, still can
not be compensated. In order to correct the fringe jump due to the discharge of NBI, the AFJC circuit is
improving to use the high phase resolution (1/1000 fringe) phase counter for the CO2 laser interferometer on
LHD as well.
Corresponding Author:
YASUHIKO. ITO
itoy@LHD.nifs.ac.jp
Oroshi-cho, Toki, GIFU, 509-5292, Japan
149
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-208
DATA ACQUISITION UPGRADE IN THE RFX EXPERIMENT
MANDUCHI GABRIELE, BARANA OLIVIERO, LUCHETTA ADRIANO, TALIERCIO CESARE
The Control and Data Acquisition system of RFX has been completely renewed and is currently under recommissioning. Most data acquisition is now performed by means of CompactPCI devices supervised by
Pentium PCs running Linux. Real-time control systems, implemented using VME and PowerPCs running
VxWorks, produce also data that are then acquired by the data acquisition system. The older CAMAC systems
have only been retained for existing diagnostics. New diagnostic systems will employ either CompactPCI data
acquisition or custom solutions, usually running under Windows, due to the fact that the drivers for the used
devices are normally available for this platform. Despite the variety of hardware and software platforms
involved in data acquisition, the same software package is used for all components, thus providing a uniform
view of the system. Such functionality is provided by the MDSplus software. MDSplus is now available for a
variety of platforms, and includes several Java components that are platform-independent. While data
organization is mostly centralized in RFX, i.e. the pulse database components are located on at most two
machines, the control and data acquisition tasks are distributed, being carried out by all the supervisory CPUs
of the CompactPCI crates, as well as by the CPUs involved in the PC-based diagnostics. This means that
during the shot sequence tens of different tasks running on different machines need to synchronize. This is
achieved by a dedicated Java component of MDSplus, which takes the required dispatching information from
the pulse file. Communication among system components is achieved using mdsip, a protocol built over
TCP/IP, which provides the “glue” that is tying MDSplus components together. A typical operating scenario
for RFX involves 15 CompactPCI crates (with 14 slots), 7 to 9 VME crates, and 2 to 5 Windows PC for
diagnostic supervision. Moreover, 15 to 20 Windows PCs are used for waveform display and graphical user
interface. Despite the large number of components, the system proved very reliable. The new configuration has
in fact been used in a set of tests for an ITER component, producing more than one thousand of shots. This is
mainly due to the robustness of the communication layer and, more generally, to the high quality of the
MDSPlus tools, a consequence of the fact that the MDSplus architecture relies on the experience gained in
years of operation on several different experiments.
Corresponding Author:
MANDUCHI GABRIELE
gabriele.manduchi@igi.cnr.it
corso stati Uniti, 4 35127 Padova - Italy
150
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-220
THERMAL DETECTOR FOR THE LOST ALPHA PARTICLE
MEASUREMENTS
ANDREY ALEKSEYEV, I.N. RASTJAGAEV 1 L.N. BUTVINA 2 V.A. DRAVIN 3
1- TRINITI, Troitsk, Moscow reg. 142190 Russia 2- General Physics Institute of the Russian Academy of Science, 38 Vavilov
st. Moscow, 117942 Russia 3- Lebedev Physical Institute of the Russian Academy of Science, 53 Leninsky pr. Moscow
119991 Russia
Recently a multi-foil thermal detector (MFTD) has been proposed for the lost alpha particle spectra
measurements1. The general idea is to measure the power deposited by a particle flux in a number of
successive ultra-thin foils. Detailed analysis shows that approximately 21 foils in the stack are required to
obtain the desirable 250 keV resolution in the total 0…3.5 MeV energy range2. In the current work we have
tested more reliable and simplified design, which comprises a number of thermal bolometers with proper frontend absorbing foil windows providing different low-energy cut-off limits for the incident particles. The
approach is similar to the well-known soft X-ray filter technique for the measurement of plasma electron
temperature. Following it, an effective temperature or a “color” of the lost alpha particles could be measured
even with a limited number of the detectors. An infrared fiber optic bolometer was developed for the thermal
sensing of the incident power. Free-standing diamond-like carbon (DLC) and chemical vapor deposited (CVD)
diamond films were used for the front-end filters. The prototype device was tested and calibrated at the He ion
accelerator in 30…700 keV energy range. 1. A.G. Alekseyev, D.S. Darrow, et al., "Nanoscale Thickness CVD
Diamond Membrane Detector for Energetic Particles Spectra and Profile Measurements", Proc. of GermanPolish Conf. on Plasma Diagnostics for Fusion & Applications, Greifswald, Germany, Sept. 4-6, 2002, B11. 2.
A.G. Alekseyev, D.V. Portnov, F.E. Cecil “Multi-foil Stack Thermal Detector for Lost Alpha Particle
Measurements”, 30th EPS Conf. on Controlled Fusion & Plasma Phys., July 7-11, 2003, St.Petersburg, Russia.
Corresponding Author:
ANDREY ALEKSEYEV
alexag@triniti.ru
TRINITI, Troitsk, Moscow reg. 142190 Russia
151
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-226
ITER RELEVANT DEVELOPMENTS OF NEUTRON DIAGNOSTICS
DURING JET TRACE TRITIUM CAMPAIGN
BERTALOT LUCIANO, J.M. ADAMS2, M. ANGELONE1, S. CONROY3, B. ESPOSITO1, Y. KASCHUCK4, D.
MAROCCO1, A. MURARI5, M. PILLON1, S. POPOVICHEV2, M. REGINATTO6, M.RIVA1, H.
SCHUHMACHER6, D. STORK2,A. ZIMBAL6 AND THE JET EFDA CONTRIBUTORS**
1Ass. Euratom-ENEA, Frascati, Italy 2Ass. Euratom-UKAEA ,Culham, UK 3EURATOM-VR Ass., Uppsala, Sweden
4TRINITI, Troitsk, Russian Federation 5Consorzio RFX – Ass. Euratom, Padova, Italy 6Physikalisch-Technische
Bundesanstalt,Braunschweig, Germany
During the JET Trace Tritium campaign a few new neutron diagnostic systems were deployed under different
plasma scenarios to provide information on the total neutron emission and its spatial and energy distribution.
The 14 MeV neutron yield was measured with one Chemical Vapour Deposited (CVD) diamond detector,
which is more resilient to neutron damage. Comparison with the JET 14 MeV monitors (Si diodes) illustrates
the good performance of the CVD device and provides an excellent correlation between the two 14 MeV yield
measurement systems. Key information on tritium transport and the behaviour of fast particles in the plasma
were obtained from the spatially and temporally measurements of neutron emission by means of the Upgraded
Neutron Profile Cameras which also provide an independent measure of the total yield. Particular attention
was paid to the operational stability of this diagnostic system. Spatial asymmetries in the neutron emission
were observed which is evidence for the influence of fast particles on the plasma. With regard to the energy
distribution of the neutron emission, a neutron spectrometer was installed based on a liquid organic scintillator
and n-g pulse shape discrimination. The results show that such systems can operate in real fusion experiments
as compact broadband neutron (from 1.5 MeV up to 20 MeV) and gamma ( from 0.3 MeV up to 10 MeV)
spectrometers with good energy resolution. Application of the digital pulse shape discrimination (DPSD)
technique with fast transient recorder cards has been successfully carried out. The main advantages of DPSD
are in enabling high count rate operation into the MHz range, and in the potential for post-experiment data
(re)processing. This new technique can be used for neutron emission counting as well for simultaneous
neutron and gamma spectroscopy The experience gained at JET indicates that these neutron measurement
systems are suitable for large fusion devices such as JET-EP and ITER where fusion neutron diagnostics will
play an increasingly important role. **See annex of J. Pamela et al., Fusion Energy 2002 (Proc. 19th Int. Conf.
Lyon, 2002), IAEA, Vienna
Corresponding Author:
BERTALOT LUCIANO
bertalot@frascati.enea.it
Associazione Euratom-ENEA sulla Fusione, C.R. Frascati, C.P. 65, Frascati, I-00044, Roma, Italy
152
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-230
OPTICAL AND ELECTRICAL DEGRADATION OF HYDROGEN
IMPLANTED KS-4V QUARTZ GLASS
S.M. GONZÁLEZ, A. MOROÑO, AND E.R. HODGSON
Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain
The light detection and ranging (LIDAR) diagnostic systems for ITER will employ high power lasers pulses
which must pass through highly transparent windows. If the laser intensity is too intense the window material
will suffer dielectric breakdown and as a consequence the window may break. The window material will be
subjected to neutron and gamma irradiation, and additionally the vacuum face of the window may be subjected
to a bombardment by low energy (eV to keV) ions and neutral particles. These low energy particles will
deposit most of their energy at or very near the surface and hence the local damage and/or degradation of the
physical properties at the vacuum surface could be very high. In particular degradation of the optical
transmission and electrical resistivity of windows are important issues for LIDAR diagnostic systems. One
candidate material for windows is KS-4V quartz glass. In the work to be presented the optical and electrical
degradation of this material implanted with protons has been addressed. KS-4V 16 mm diameter, 1 mm thick
disc samples, were implanted at 25 C with hydrogen ions (protons) of energies between 30 and 55 keV, 1
microamp/cm2, up to a dose of 10+16 ions/cm2. After implantation the optical absorption from 195 to 3000
nm was measured, and then the samples were mounted in a system which permitted one to measure the surface
and volume electrical conductivities in high vacuum (10-6 torr) at temperatures between 20 and 450 C. Proton
implantation of KS-4V quartz glass modified the optical transmission, producing a monotonic increase in the
absorption extending from the IR to the UV region. The measured absorption at 800 nm implies that 30 % of
the laser power would be absorbed in a surface region of about 1 µm. In addition the electrical conductivity of
the material over the implanted area severely increases by many orders of magnitude. These two effects
dramatically enhance the risk of laser breakdown if the material is to be used as a LIDAR window. The
enhanced surface electrical conductivity increases with increasing temperature, and the observed behaviour
suggests the formation of a semiconducting material layer. This type of surface electrical degradation should
be studied in other insulators to be used in ITER not only in diagnostics but also in heating systems where
electrical insulation is of concern.
Corresponding Author:
S.M. GONZÁLEZ
sm.gonzalez@ciemat.es
Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain
153
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-234
RECONSTRUCTION CAPABILITY OF JET MAGNETIC SENSORS
CENEDESE ANGELO, RAFFAELE ALBANESE (1) GIOVANNI ARTASERSE (1) MASSIMILIANO MATTEI (1)
FILIPPO SARTORI (2)
(1) Assoc. EURATOM-ENEA-CREATE, DIMET, Univ. Mediterranea di Reggio Calabria, Via Graziella, Loc. Feo di Vito, I89060 Reggio Calabria, Italy (2) Euratom/UKAEA Fusion Assoc., Culham Science Centre, Abingdon, Oxon, OX14 3DB, UK
During the 2004 shutdown, a new set of magnetic sensors will be installed in JET, designed so as to upgrade
and in part to replace the existing diagnostic system. For this reason, as far as the reconstruction capability is
involved, it has been fundamental for the sensor definition to quantify beforehand how much the magnetic
enhancement will -in principle- augment the measurability of the plasma shape, and therefore extend the JET
operating space. To discern between an increased redundancy in the measure and new information brought in
by the sensors, a model based statistical analysis resorting to the correlation function among the magnetic
measurements has been carried out. To perform this study, experimental and simulated databases have been
constructed, spanning the whole variety of already achievable plasmas and the designed new JET-EP
configurations for 2005. In addition, the field reconstruction error in proximity to the plasma boundary has
been assessed using the present and the augmented sensor sets, which gives indication on the achievable
accuracy of the plasma boundary itself. As a matter of fact, besides representing a cross-validating and
consistent study on JET magnetics, both these analyses seem to confirm that an enhanced reconstruction can
be obtained, with a noise amplitude reduction in localised parts of the boundary (up to 50% on the outboard).
The methodology sets also the guidelines for the development of software tools useful for the experimental
commissioning and measure validation of the sensors.
Corresponding Author:
CENEDESE ANGELO
cenedese@igi.cnr.it
Consorzio RFX Assoc. EURATOM ENEA sulla Fusione, Corso Stati Uniti, 4, I-35127 Padova Italy
154
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-235
QUENCH DETECTION & DATA ACQUISITION SYSTEM FOR SST-1
SUPERCONDUCTING MAGNETS
A. N. SHARMA, C.J. HANSALIA, Y. YEOLE, G. BANSAL, S. PRADHAN, & Y.C. SAXENA
INSTITUTE FOR PLASMA RESEARCH, GANDHINAGAR
Superconducting magnet system of Steady State Superconducting Tokamak (SST-1) shall be operating in a
very noisy environment. Presence of high inductive voltages in the magnets during off-normal events like
VDE, plasma disruption, and PF magnet ramp ups etc, has made quench detection and data acquisition a
challenging task. A hybrid of analog electronic circuits and software controlled data acquisition system has
been developed and tested to safeguard the magnets. This paper will describe the electronic hardware circuits
developed for signal conditioning, high voltage suppression, fail proof quench detection and for noise
elimination algorithms and their testing. The SST-1 Superconducting magnets will have large number of
sensors like voltage taps, Venturi flow meter, strain gages, hall probes, pressure sensors, temperature sensors,
and displacement transducers. A real time data acquisition system has been designed using VMEbus for
monitoring and storing signals from all these sensors and initiating control action in case of off-normal events.
The paper will also describe the configuration of the data acquisition system with emphasis on hardware used
and the software developed for it.
Corresponding Author:
A. N. SHARMA
ansharma@ipr.res.in
INSTITUTE FOR PLASMA RESEARCH, BHAT, GANDHINAGAR - 382428, (GUJARAT) INDIA
155
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-248
LASER DAMAGE OF KU-1 SILICA GLASS COVERED WITH
HYDROCARBON FILM
KONSTANTIN VUKOLOV, DORIAN ORLINSKI (1) ALEXANDER GORBUNOV (2) NIKOLAY KLASSEN (2)
(1)RRC Kurchatov Institute, Kurchatov Sq.46, 123182 Moscow, Russia (2)Institute of Solide State Physics, 142432
Chernogolovka, Moscow region, Russia
Results are given of the experimental investigation of a laser damage of diagnostic windows covered with
hydrocarbon films under the pulse laser irradiation (YAG laser, t=10 ns, f=7-33 Hz and an energy up to 0.4
J/pulse). The optically polished samples (diameters of 10 and 16 mm, thickness 10 mm) of KU-1 silica glass
which is planned to be used in ITER were tested. One side of the samples was covered with thin hydrocarbon
films (CH and CD with different content of deuterium) of different thickness (10, 50 and 100 nm). Laser beam
was introduced through the clean side of widow sample and was focused on the filmed surface with a spot of
0.2 mm in diameter for laser damage threshold measurements and of ~1 mm in diameter for measurements of
hydrocarbon films ablation (cleaning threshold). The experiments were done in atmosphere. The cleaning
threshold was determined by monitoring of film surface in scattered or reflected light and laser damage
threshold – on characteristic click and blue light appearance. Measured cleaning thresholds of hydrocarbon
films were 0.1-0.3 J/cm2 depending on the deuterium content and to a lesser degree on the film thickness. The
laser damage thresholds were about of 50-60 J/cm2 for KU-1 windows with films on the surface and about of
80 J/cm2 for clean windows. An error of these measurements was 15-20%. Probably same laser damage
thresholds for the samples with and without films mean that the film evaporation takes place during the first
laser pulse. The expected interaction of film material and glass was not observed. However graphite erosion
products will be always presented into ITER vessel and the window surface may be contaminated with
hydrocarbons penetration into the glass under laser radiation. This year it is supposed to repeat the experiments
in a vacuum chamber with simulation of the ITER vacuum conditions.
Corresponding Author:
KONSTANTIN VUKOLOV
vukolov@nfi.kiae.ru
RRC Kurchatov Institute, Kurchatov Sq.46, 123182 Moscow, Russia
156
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-250
WIDE AREA DATA REPLICATION IN AN ITER-RELEVANT DATA
ENVIRONMENT
IANNONE FRANCESCO, GIOVANNI BRACCO (1) CRISTINA CENTIOLI (2) SILVIA ECCHER (3) MAURIZIO
PANELLA (2) MAURIZIO STEFFE' (4) VINCENZO VITALE (2)
(1) ENEA-INFO, Via Enrico Fermi 45, 00044 Frascati (RM) Italy (2) same as main author (3) CASPUR Via dei Tizii, 6b
00185 Roma Italy (4) ENEA - Centro ricerche della Casaccia Via Anguillarese, 301 00060 S. Maria di Galeria (RM) Italy
The next generation of tokamak experiments will require a new way of approaching data sharing issues among
fusion organizations. In the fusion community, many researchers at the different sites worldwide could analyze
data produced by ITER wherever it will be built. Usually in such large size experiments an approach is the
efficient availability of the data near to the location where the computational resources are available. This new
approach should go beyond the site-centric model mainly devoted to granting access exclusively to
experimental data retained in the device sites. To this aim, we propose a new data replication architecture
relying on a wide area network, based on master/slave model and synchronization techniques producing
mirrored data sites. In this architecture, data replication will affect large databases (TB) as well as large unixlike file systems, using open source based software components, namely MySQL as database management
system, and rsync and bbftp for data transfer. A testbed has been set up to evaluate the performance of the
software components underlying the proposed architecture. The testbed hardware layout deploies a cluster of
four Dual-Xeon Supermicro each with a raid array of 1 Terabytes. High performance network lines (1 Gbit
over 400 Km) will provide to test the components on wide area network. The results obtained will be
thoroughly discussed.
Corresponding Author:
IANNONE FRANCESCO
iannone@frascati.enea.it
ENEA - Centro Ricerche Frascati, Via Enrico Fermi, 45 I-00044 Frascati (RM) Italy
157
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-251
ADVANCES IN REMOTE PARTICIPATION FOR FUSION
EXPERIMENTS*
SCHISSEL, D.P., V. SCHMIDT (1), J.W. FARTHING (2)
(1) Consorzio RFX, Associazione Euratom-ENEA sulla fusione, Padova, Italy (2) Association Euratom-UKAEA, Culham
Science Centre, Abingdon, United Kingdom
Selected also for oral presentation
O2B-D-251
Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in
collaborations between experimental sites and laboratories worldwide. In the US, fusion experimental research
is centered on three large facilities involving over 1000 researchers covering 37 states. In the EU, fusion
research is coordinated by EFDA, encompassing some 25 laboratories and several major facilities, including
JET. Collaborative research within each group, combined with collaboration between the two groups is
presenting new and unique challenges in the field of remote participation technology. These challenges are
being addressed by the creation and deployment of advanced collaborative software and hardware tools. Grid
computing, the secure integration of computer systems over high-speed networks to provide on-demand access
to data analysis capabilities and related functions, is being deployed as an alternative to traditional resource
sharing among institutions. Utilizing public-key based security that is recognized between the EU and US, the
TRANSP analysis code is running on one cluster yet is securely available worldwide. This analysis service
includes secure remote data access as well as advanced web-based monitoring capabilities. Traditional audio
teleconferencing is being augmented by more advanced capabilities including videoconferencing, instant
messaging, presentation sharing, applications sharing, tiled display walls, and the virtual-presence capabilities
of the AccessGrid, with its potential for remote control room presence. With these advances, remote real-time
experimental participation has begun including several remotely led JET experiment sessions where the lead
scientists were in other laboratories in the EU or the US. Work continues to focus on reducing the variety of
remote participation methods, on improving interoperability between the different approaches, on ease of use,
and on improved security. The collaborative technology being deployed is scalable to fusion research beyond
the present programs, in particular to the ITER experiment that will require extensive collaboration capability
worldwide. This paper will compare approaches, review the present state-of-the-art in remote participation
capability, and identify areas of work required for the success of future large-scale experiments. *Work
supported by U.S. Department of Energy under DE-FC02-01ER25455 and by the European Fusion
Development Agreement (EFDA)
Corresponding Author:
SCHISSEL, D.P.
schissel@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
158
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-258
SOFT COMPUTING AND CHAOS TEORY FOR ANTICIPATION OF
DISRUPTION IN TOKAMAK REACTORS
FRANCESCO CARLO MORABITO, DOMENICO COSTANTINO MARIO VERSACI
Università di Reggio Calabria. Facoltà di Ingegneria - DIMET
Disruption represents a transfer of energy of the plasma to the surrounding mechanical structures. During the
sudden loss of confinement, the energy content collapse in an uncontrollable way, generating mechanical
forces and heat loads which threaten the structural integrity of surrounding structures and vacuum vessel
components. It is thus of primary importance to design an alarm system for detecting he onset of a disruption.
Neural Network models have been proposed in the recent literature as forecasting systems, with the aim of
predicting the occurrence of disruptions sufficiently far in advance for protecting procedures to be switched
on. Then, it is necessary to check the incoming of disruption by means of a suitable time window of prediction
to take into account control actions. The system is subjected to the fusion reactions and it is really complex
according to Chaos Theory. In this paper, chaotic analysis and soft computing approach are exploited to
predict the incoming of disruptions. In particular, we propose a neuro-fuzzy approach cooperated with chaos
theory for the prediction problem above mentioned. By means of chaos theory, it is possible to compute a
suitable time window for prediction problem. The aim is to establish the presence of chaos and its degree in
the system under study in order to extract a time window for prediction. Power Spectrum Density is computed
to determine the presence of chaos in data. Then, states space determination is carried out in order to extract
the emdedding dimension (d), that establishes the number of coordinates in which each point of states space is
represented, and the time lag (t), that represents an integer multiple of sampling period exploited to reconstruct
the time series. The goal of this process is to reconstruct the strange attractor that conserves the topological
properties of the system. By means of d and t it is possible to compute Large Lyapunov Exponent (LLE) which
measures the divergence of nearby trajectories. LLE can be exploited to compute the Horizon of Prediction
(HOP) which represents the short-term predictability. In addition, d and t have been exploited to make a non
linear predictive model for our problem. In particular, we propose the use of Neural Networks and Fuzzy
Inference Systems to predict the incoming of disruption starting from chaotic parameters. The obtained results
have shown a good prediction with respect to time window computed by HOP.
Corresponding Author:
FRANCESCO CARLO MORABITO
morabito@unirc.it
Università di Reggio Calabria. Facoltà di Ingegneria - DIMET
159
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-263
ADSORPTION IN INSULATOR MATERIALS ENHANCED BY D
IMPLANTATION
A. IBARRA, A. CLIMENT-FONT (1.2) A. MUÑOZ-MARTÍN (2)
(1) Dpto. Fisica Aplicada, Uni. Autónoma Madrid, 28049 Madrid, Spain (2) CMAM, Uni. Autónoma Madrid 28049 Madrid,
Spain
Insulator materials are critical components of different heating and current drive systems as well as for many
diagnostics to be used in the ITER machine. Insulator materials are used as optical and dielectric windows and
as electric insulator in cables, feedthroughs and connectors. In many of these applications, the insulator surface
is exposed to a “dirty” gas phase composed of DT gas and plasma with hydrocarbons, stainless steel, W and
Be particles, produced by the erosion of the first wall materials by the fusion plasma, combined with the
presence of an intense neutron and gamma radiation field and high electric and magnetic fields. These particles
can be deposited on the surface of the insulator giving rise to degradation of their properties, like, for example,
the presence of hot spots in radiofrequency windows or the increase of electrical conductivity. No clear
information is actually available on the characteristics and deposition rate that can be actually expected for
ITER applications, neither on the possible effect of the radiation or electromagnetic fields on these properties.
In this work, a number of different insulator candidate materials (Al2O3, SiO2, diamond) are implanted at
room temperature with low energy (several keV) D and H ions in order to qualitatively simulate some of the
DT gas effects. The implantation effects are characterized using optical absorption and Elastic Recoil
Detection (ERD) techniques. It is observed an increase of the optical absorption due to the implantation that
can be related to an increase of the surface scattering. In parallel it can be observed an increase in the C and H
adsorbed at the surface. These effects are observed for all the studied materials and suggest that the
implantation degrades the surface characteristics increasing its adsorption capability. This can induce
important effects on the long term behaviour of insulator materials for fusion. The implantation energy and
dose dependence of the measured behaviour will be discussed.
Corresponding Author:
A. IBARRA
angel.ibarra@ciemat.es
Euratom / CIEMAT Fusion Association, 28040 Madrid, Spain
160
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-270
RADIATION ENHANCED DEGRADATION OF SIO OVERCOATED
ALUMINIUM MIRRORS
E.R. HODGSON, T. HERNANDEZ, AND A. MOROÑO
Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain
High quality mirrors for the optical UV - visible - IR range will be required in ITER for both remote handling
applications and diagnostic systems. The reflectivity of the mirror surface can be degraded by many different
mechanisms ranging from simple contamination, to sputtering erosion due to particle bombardment, and more
complex enhanced radiation effects which may modify the surface structure. The commercially available high
quality mirrors being considered for ITER applications consist of a thin evaporated aluminium layer on a solid
glass substrate. To protect the delicate aluminium surface the mirrors are covered (overcoated) with a
controlled layer of transparent protective material such as SiO, of adequate thickness to obtain optimum
optical constructive interference in a given wavelength range, usually the visible region. The work to be
presented describes experiments carried out in order to study radiation enhanced degradation of such high
quality mirrors, in particularly the importance of the surrounding operational environment. Tests were made on
Coherent research quality SiO overcoated Al on Pyrex glass mirrors. The irradiations were performed in a
sample chamber mounted in the beam line of a 2 MeV Van de Graaff accelerator. The chamber permits
irradiations perpendicular to the reflecting surface to be performed in different environments. In this way
mirror samples were irradiated at 25 C in dry air, nitrogen, and high vacuum (10-6 mbar) with a 1.8 MeV
electrons at 150 Gy/s up to a dose of 10 MGy. Reflectivity measurements from 250 to 2500 nm were made at
different irradiation doses and compared with the reflectivity before irradiation. The mirrors irradiated in
nitrogen and vacuum did not show any measurable changes in reflectivity. However mirrors irradiated in dry
air exhibited a marked reflectivity decrease in the visible (violet/blue region) and particularly in the UV range.
This is believed to be due to radiation enhanced oxidation of the SiO protective layer. As the SiO layer
oxidizes both thickness and refractive index of the layer are modified hence degrading the optimised optical
interference process.
Corresponding Author:
E.R. HODGSON
hodgson@ciemat.es
Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain
161
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-272
THE NEW MEASUREMENT MONITORING SYSTEM ON FTU
BERTOCCHI ALFREDO, SALVATORE PODDA (1) VINCENZO VITALE (1)
(1) Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy
Needs concerning the possibility for customers to easily plot a wide range of FTU subplants measurements,
particularly the named “slow measurements”, have lead to redesign the related acquisition system. Opto22
modules – which use Ethernet as fieldbus – were adopted to substitute old PLC devices. An appropriate
MySQL database is updated by a C++ Opto22 driver and a CORBA server, running on the same machine
hosting the MySQL server, allows the database access to any remote CORBA client by means of some specific
methods. In the previous architecture the new signal addition meant a lot of configuration work based on the
use of proprietary and dedicated tools. Moreover the monitoring system constrained to use a non user-friendly
graphical interface based on a commercial software package with a strong platform dependence (Digital Unix).
Vice versa the new situation allows to overcome these limitations: it makes possible to easily configure
Opto22 modules and MySQL database within a browser while data management and visualization are
achieved by using a graphical interface developed in Java, which yields these operations completely platform
independent. In addition the CORBA server introduces the following advantages: 1. a hardware independence,
giving maximum flexibility about the choice of platforms and devices as system components, 2. both network
and programming language transparency. A remarkable aspect looks out for the use of packages totally free
software. This paper will present the new system architecture, last results and future developments.
Corresponding Author:
BERTOCCHI ALFREDO
bertocchi@frascati.enea.it
Associazione Euratom/ENEA sulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati (Roma), Italy
162
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-281
FIRST RESULTS OF MINIMUM FISHER REGULARISATION AS
UNFOLDING METHOD FOR JET NE213 LIQUID SCINTILLATOR
NEUTRON SPECTROMETRY
JAN MLYNAR, JOHN M ADAMS (1) LUCIANO BERTALOT (2) SEAN CONROY (3)
(1)Association EURATOM-UKAEA Fusion, Culham Science Centre, Abingdon, Oxfordshire, OX143DB, UK (2)Association
EURATOM-ENEA sulla Fusione, Frascati, C.P. 65, 00044-Frascati, Italy (3)Association EURATOM-VR, Uppsala
University, SE-751 05 Uppsala, Sweden
At JET, the NE213 liquid scintillator is being validated as a diagnostic tool for spectral measurements of
neutrons emitted from the plasma. The main goals of the diagnostic are to measure characteristics of 2.5 MeV
(D-D fusion) and 14 MeV (D-T fusion) spectra, and to evaluate the neutron fraction due to ICRH fast ions.
The relation between acquired pulse height spectrum and source neutron spectrum is not straightforward. The
response function matrix of the scintillator is based on theoretical prediction and accelerator calibration. This
matrix, when multiplied by the (unknown) neutron spectrum, gives the (measured) pulse height spectrum. The
unfolding process thus consists of finding an inverted solution. This is an ill-conditioned problem which,
without further constraints, leads to unrealistic amplifications of systematic errors and noise. The Minimum
Fisher Regularisation (MFR) has been applied in two-dimensional tomography of SXR and bolometric
measurements at the TCV tokamak. It is a direct inversion method, which constrains the object function
smoothness, providing robust fits rapidly. At JET, the Maximum Entropy Method has been applied to neutron
spectra unfolding, with occasional ambiguous results. The MFR, which presents a completely different
inversion algorithm, has been implemented as an independent and transparent tool to validate the JET neutron
spectra measured with the NE213 liquid scintillators. The adapted MFR algorithm was first thoroughly tested
on phantom spectra. Following the encouraging results, real pulse height spectra from the JET NE213 neutron
detector were successfully analysed from D-D, Trace Tritium and ICRH fast ions experiments. These first
applications of MFR tend to confirm independent JET results, e.g. presence of high energy (approx. 4 MeV)
alpha particles in experiments studying their acceleration by 3rd harmonics ICRH waves.
Corresponding Author:
JAN MLYNAR
jan.mlynar@jet.efda.org
EFDA-JET CSU, Culham Science Centre, OX14 3DB Abingdon, UK / Association EURATOM-IPP.CR,
Institute of Plasma Physics AS CR, CZ-182 21 Prague 8, Czech Republic
163
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-282
PRESENT STATUS OF THE TJ-II REMOTE PARTICIPATION SYSTEM
VEGA, JESUS, SÁNCHEZ E. (1) LÓPEZ A. (1) PORTAS A. (1) OCHANDO M. (1) ASCASÍBAR E. (1)
MOLLINEDO A. (2) MUÑOZ J. A. (2) SÁNCHEZ A. (2) RUIZ M. (3) BARRERA E. (3) LÓPEZ S. (3) CASTRO R.
(4) LÓPEZ D. (4)
(1) Asociación EURATOM/CIEMAT para Fusión. Madrid. Spain. (2) CIEMAT. Computing Center. Madrid. Spain. (3) UPM.
Dpto. de Sistemas Electrónicos y de Control. Madrid. Spain. (4) Red.es. Madrid. Spain
The TJ-II remote participation system (RPS) has been designed to extend to INTERNET the present working
capabilities provided in the TJ-II local environment, i.e., tracking the TJ-II operation, monitoring/programming
data acquisition and control systems, and accessing databases. Critical aspects considered in this development
arose as a consequence of three factors: - Growth capabilities. The system architecture had to be flexible
enough to permit new requirements to be incorporated at any point in time. - Open system. Local TJ-II group
platforms cannot be imposed on remote participants. - Working environment. Firstly, security is “a must” for
systems exposed to the INTERNET. Secondly, software distribution and version control can be a major
problem in local area networks with the result that any negative effects could be amplified in INTERNET.
Thirdly, the unavoidable administration tasks of the system cannot be forgotten. After taking into account the
above constraints and the capabilities to be provided, the TJ-II remote participation system was based on web
technologies. A web server acts as a communication front-end between remote participants and local TJ-II
elements. The remote participation system is based on Java Technology because of its open character, security
properties and technological maturity. From the server side, network services are provided by means of
resources supplied by JSP pages. The client part makes use of web browsers and ad-hoc Java applications.
Regarding software deployment, use was made of solutions based on the Java Network Launching Protocol.
The operation requires the use of a distributed authentication and authorization system for filtering user
queries. This development employs the PAPI System. At present, approximately 1000 digitization channels
can be managed from the TJ-II RPS. Furthermore, processing software based on a 4GL language (LabView)
can be downloaded to multiprocessor data acquisition systems. Also, 15 control systems are available from the
RPS. Databases and the TJ-II operation logbook are available (the system even allows for the physicist in
charge of operation to be in a remote location). Audio/video resources allow on-line access to the TJ-II control
room and at present four Spanish universities make use of the TJ-II remote participation system capabilities for
joint collaborations: these are the UPM, UNED, UCM and UPC.
Corresponding Author:
VEGA, JESUS
jesus.vega@ciemat.es
ASOCIACION EURATOM/CIEMAT PARA FUSION. Avda. Complutense, 22. 28040 Madrid (SPAIN)
164
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-347
APD DETECTOR ELECTRONICS AND PXI BASED DATA
ACQUISITION SYSTEM FOR SST-1 THOMSON SCATTERING
DIAGNOSTICS
CHAVDA CHHAYA, T.ARUNA,K.PATEL,Y.C. SAXENA,AJAIKUMAR
Institute for Plasma Reserach, Near Indira bridge, Bhat,Gandhinagar, Gujarat,382 428 India
An electronic system with optimum signal conditioning has been designed and tested for a thermoelectrically
cooled Si-avalanche photodiode (APD) to measure the Thomson scattered spectrum from the SST-1 tokamak
plasma. The electronic system consists of a current-feedback preamplifier and the read-out unit. The read-out
system has preamplifiers with a fast (50 MHz) and slow (1 MHz) output. The fast output is first delayed by
100ns and then the delayed and non-delayed output are processed by a differential amplifier to subtract low
frequency background light component. The slow output signal is further amplified with a computer control
variable gain amplifier for the purpose of error measurement and calibration. To avoid ground loops and other
pickups, fast signal output is digitized by in-house developed charge to time converter in the same electronics
read out system. The fast time data from detector electronics system is converted to digital data by time to
digital converter using TDC from LeCroy. The prototype charge to time converter is tested using CAMAC
based precision charge/time generator module with charge varies from 10pC to 600pC with a gate of 20nsec.
PXI based system transfer the digital data to PC for further analysis. Multipoint Thomson scattering system
needs large number of charge coupled digitizer channels. For reducing the number of charge ADCs, a scheme
is worked out to multiplex the charge signals. To test the concept a prototype of 4-channel multipexer system
is developed and tested for simultaneous sampling with standard CAMAC based modules.
Corresponding Author:
CHAVDA CHHAYA
chhaya@ipr.res.in
Institute for Plasma Research
165
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-366
REAL TIME MEASUREMENT AND CONTROL AT JET - DIAGNOSTIC
SYSTEMS
MURARI, ANDREA (3), ROBERT FELTON(1) JOFFRIN, EMMANUEL(2)
(1) Euratom/UKAEA Fusion Association, Culham Science Centre Abingdon OX14 3DB UK (2) Association EURATOM-CEA,
CEA Cadarache 13108 Saint-Paul-lez-Durance France
To meet the requirements of the scientific programme, the EFDA JET Real Time Measurement and Control
Project has developed an integrated set of real-time plasma measurements, experiment control, and
communication facilities. Traditional experiments collected instrument data during the plasma pulse and
calculated physics data after the pulse. The challenge for continuous tokamak operation is to calculate the
physics data in real-time, keeping up with the evolution of the plasma. In JET, many Plasma Diagnostics have
been augmented with extra data acquisition and signal-processing systems so that they can both capture
instrument data for conventional post-pulse analysis and calculate calibrated, validated physics results in realtime. During the pulse, the systems send sampled data sets into a network, which distributes the data to several
destinations. These may do further analysis integrating data from several measurements or may control the
plasma scenario by heating or fuelling. The simplest real-time Diagnostic systems apply scale factors to the
signals, as with the Electron Cyclotron Emission Diagnostic’s 96 tuned radiometer channels, giving the
electron temperature profile. In various Spectroscopy Diagnostics, spectral features are least-squares-fitted to
measured spectra from several lines of site, within 50 ms. Ion temperatures and rotation speed can be
calculated from the line widths and shifts. For Diagnostics using modulation techniques, the systems
implement digital-signal processing phase trackers, lock-in amplifiers and filters. The interferometer samples
15 channels at 400 kHz for 30 s, i.e. 6 million samples/sec ! Diagnostics have specific lines of sight, spatial
channels and various sampling rates. The Heating/Fuelling systems have relatively coarse spatial localisation.
Analysis systems have been developed to integrate the basic physics data into smaller sets of controllable
parameters on a common geometry e.g. temperature, density and safety factor profiles with values at 10 points
of normalised radius. The EFDA Real Time project is essential groundwork for future reactors such as ITER,
and has successfully involved many scientific and technical staff from several institutions. The facility is now
frequently used in experiments. This work has been conducted under the European Fusion Development
Agreement and is partly funded by Euratom and the UK Engineering and Physical Sciences Research Council.
Corresponding Author:
MURARI, ANDREA (3)
amura@jet.uk
(3) Euratom/UKAEA Fusion Association, Culham Science Centre Abingdon OX14 3DB UK
166
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-370
OPTICAL FIBERS FOR PLASMA DIAGNOSTICS UNDER GAMMARAY AND UV IRRADIATION
SPOREA DAN, ADELINA SPOREA(1) BOGDAN CONSTANTINESCU(2)
(1)National Institute for Lasers, Plasma and Radiation Physics, Atomistilor St.409, 76900 Magurele Romania. (2) National
Institute for Physics and Nuclear Engineering, Atomistilor St.409,76900 Magurele Romania
Selected also for oral presentation
O2B-D-370
The use of optical fibers under irradiation conditions (gamma, UV, neutron) is of interest, if we consider the
role they can play in diagnostics, remote handling and control, in future fusion installations (i.e. ITER,
DEMO). The purpose of our investigation was to evaluate the combined effects of various factors (UV
radiation, gamma irradiation, temperature stress) on the optical transmission of optical fibers, which can be
candidates for the development of light guides. We used large core diameter optical fibers (400 microns),
having an enhanced UV transmission, and resisting to high temperatures. The optical fibers employed are
commercially available products. They were subjected to different treatments: - gamma irradiation and
temperature cycling; - UV exposure and temperature cycling; UV, - gamma and temperature cycling. The UV
exposure was done with a stabilized CW operation, continuum spectrum deuterium lamp, which was also used
for the measurement of the spectral absorption of the optical fiber. Gamma irradiation was done in 250 krad
steps, up to a total dose of 1500 krad. All the measurements were carried out off-line, before and after each
treatment, with the set-up based on a multi-channel optical fiber spectrometer. The measurements covered the
spectral range from 200 nm to 1000 nm. For the case we consider the base-line form the spectrum graphs as
the absorption corresponding to the non-irradiated optical fibers (a value of 0.7 units in a logarithmic scale),
our investigations indicate: - the wavelength corresponding to the UV induced peak absorption is at about 220
nm; - the UV induced absorption seems to saturate (i.e. a 2 h UV exposure induces an absorption peak of 0.9
units, a 6 h exposure one of 1.05 units, while an exposure of 10 h produces a peak of 1.08 units); - temperature
stress applied after UV exposure reduces the optical absorption in a limited range (about 0.4 - 0.6 units); - the
peak for gamma induced optical absorption is located between 230 nm and 235 nm, depending on the gamma
total dose and the temperature stress applied; - in the case of pure gamma irradiation (no UV exposure) a
temperature stress reduces the absorption peak by 0.3 unit; - any temperature treatment following a gamma
irradiation produces a decrease of the optical absorption, but the radiation effect can not be reduced
completely.
Corresponding Author:
SPOREA DAN
sporea@ifin.nipne.ro
Department of Lasers, National Institute for Lasers, Plasma and Radiation Physics, Atomistilor St. 409,
Magurele 76900 Romania
167
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-384
THE HALO CURRENT SENSOR SYSTEM FOR JET-EP
MARCUZZI DIEGO, P.SONATO(1) W.BAKER(1) P.BEAUMONT(2) T.BOLZONELLA(1) C.DAMIANI(3)
P.FIORENTIN(1) A.GUIGON(3) K.FULLARD(2) A.GOODYEAR(2) L.GRANDO(1) S.HUNTLEY(2) N.LAM(2)
A.LIOURE(3) A.LOVING(2) S.MILLS(2) S.PERUZZO(1) N.POMARO(1) V.RICCARDO(2) M.WAY(2)
(1)Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, 35127 Padova Italy (2)UKAEA/Euratom Fusion
Association, Culham Science Centre, Abingdon, Oxfordshire OX14 DB United Kingdom (3)EFDA-JET-CSU Enhancement
Department,Culham Science Centre
Vertical instability of elongated plasma is a problem theoretically predicted and experimentally observed on
many different devices. During Vertical Displacement Events (VDEs), currents flowing through plasma and
vacuum vessel are present. These currents, called Halo Currents (HCs), induce severe mechanical stresses on
the plasma facing components and in the vessel, and are a major concern for present and future fusion
experiments The need to better understand the origin, the distribution, and the scaling of HCs is one of the
critical points for any next step device like the ITER project. The new system of Halo Current Sensors (HCS)
designed for JET-EP should help evaluating HC density distribution, localization and rotation as well as
toroidal and poloidal current asymmetries, their nature and correlation with other plasma parameters. The
system will be integrated with the sensors (toroidal pick up coils, Rogowski coils and mushroom tiles
instrumented with shunts) already in operation. The new system consists of Rogowski coils and toroidal field
pick-up coils. The Rogowski coils will measure directly the current flowing through some of the tiles of the
upper dump plate. They are housed in a groove machined in the CFC tiles and are designed to collect the
current flowing through one single tile. The toroidal field pick-up coils will estimate the total poloidal HC. The
HCS system will include 4 identical mechanical structures each including: 8 Rogowski coils and 2 toroidal
pick up coils. The coil assemblies are installed at the top of the vessel close to secondary X point in 4 octants
equally spaced along the toroidal coordinate. Both sensors are coils wound around a ceramic core. They have
to withstand temperature up to 400 C, therefore the windings have to be made of mineral insulated cables; the
cable section is only 0.65 mm to have a sufficiently large effective area. The mineral insulated cables have to
be UHV sealed and must withstand a voltage up to 1000 V in DC and therefore a special termination has been
qualified. A special procedure for the UHV leak tightness qualification of the terminations has been
implemented. Special tools and procedure have been implemented and tested for the remote installation of the
four assemblies. In the paper the final design will be presented in detail, as well as the results of the tests
performed for the qualification of components and the main features of the manufacturing phases of the
complete system.
Corresponding Author:
MARCUZZI DIEGO
diego.marcuzzi@igi.cnr.it
Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, 35127 Padova, Italy
168
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-396
DEVELOPMENT OF ACTIVELY COOLED PERISCOPES FOR
DIVERTOR OBSERVATION
KOENIG, RALF, K. GROSSER(1), D. HILDEBRANDT(1), O. OGORODNIKOVA(2), C. VON SEHREN(1) AND T.
KLINGER(1)
(1)Max-Planck-Institut für Plasmaphysik, Wendelsteinstr. 1, 17491 Greifswald Germany. (2)IWV-II, FZ-Jülich GmbH, D52425 Jülich Germany
With the stellarator W7-X a step to quasi-continuous plasma operation will be made. The cooling system of
the machine is designed such that two 30 min discharges can be run per day. Right from the start of operation
10 MW of ECRH heating power will be available for quasi-continuous operation. A working group “Plasma
Facing Optical Components” has been formed which presently concentrates on the development of water
cooled shutters and windows for UV/Vis./IR periscopes which can withstand the expected maximum heat
loads of up to 50 kW/m2 which due to the predominantly short wavelength nature of the radiation emitted by
the plasma will be absorbed within the first millimeter of any window. We will report on the detailed ANSYS
calculations of the heat and stress distribution across the shutter, the bottom of the immersion tube and in
particular the windows. In the case of the windows calculations have been undertaken for a large number of
different materials which are required for the various spectral regions covered by the miscellaneous
diagnostics, so that the most suitable material for each application can easily be identified. Also the
dependence of the cooling rate on the window diameter and thickness has been studied. The calculations
suggest that CVD-diamond which might be required as a very expensive last resort for a few large windows
(dia. >150 mm) can most likely be avoided by using sapphire but that for many of the other materials like
ZnSe, ZnS, CaF2, MgF2 and quartz one will be limited to considerably smaller sizes. A vacuum test chamber,
equipped with a vacuum compatible IR heater has been build. In this chamber a low cost, easily exchangeable
window design using Helicoflex gaskets on either side of Sapphire and Quartz windows has been successfully
tested. The design was water tight and the window materials behaved roughly as predicted by the ANSYS
calculations, with sapphire, as expected, showing excellent heat removal properties. The test windows are
being blackened to ensure effective absorption of the IR radiation at the surface of the substrates on the
vacuum side of the windows and IR cameras for different wavelength regions as well as other test equipment
were installed at the periphery of the test chamber for detailed investigations of the time evolution of the radial
heat distribution across the different window materials at different power loads. These results will be
compared with the ANSYS calculations.
Corresponding Author:
KOENIG, RALF
rlk@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Wendelsteinstr. 1, 17491 Greifswald, Germany
169
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-398
DIAGNOSTICS FOR STUDYING DEPOSITION AND EROSION
PROCESSES IN JET
COAD, JOSEPH PAUL, H-G ESSER(1), J. LIKONEN(2), M MAYER(3), G NEILL, V PHILIPPS(1), M.
RUBEL(4), J VINCE AND EFDA-JET CONTRIBUTORS
(1)Forschungszentrum Jülich, D-52425, Jülich, Germany (2)VTT Processes, P.O. Box 1608, 02044 VTT, Finland (3)MaxPlanck Institut für Plasmaphysik, D-85748 Garching, Germany (4)Alfvén Laboratory, Royal Institute of Technology, 100 44
Stockholm, Sweden
Information on how and where hydrogen isotopes are trapped in JET is important in understanding the
erosion/deposition process well enough to predict performance in ITER. In the past this information has been
largely limited to the analysis of first wall components removed after complete operational campaigns. New
complementary diagnostics are being developed for installation in JET in 2004 under the JET Enhancement
Programme and Task Force Fusion Technology. A prototype quartz micro-balance (QMB) was developed for
JET to measure deposition on-line in a time-resolved manner, and has worked within the JET divertor for two
years [1]. New QMBs are being developed from this prototype for installation in 2004. These QMBs are
designed to measure deposition in the septum and outer divertor as well as the inner divertor, explore the effect
of temperature on deposition from room temperature to ~300?C, and monitor the amount of beryllium
evaporated in JET. New collector units are also being developed that have a rotation mechanism powered by
the magnetic field. Erosion/deposition will be monitored over ~3000 pulses with a time resolution of about 50
pulses. Since the units need no electrical connections, they will be fitted to the outer vessel wall in addition to
the divertor. However, the data on the collector can only be evaluated after the unit is retrieved from the
vessel. The transport processes involved in deposition in shadowed areas of JET are not understood. First
information that most of the hydrocarbon radicals involved have high sticking coefficients has come from
deposition monitors in JET [2]. More deposition monitors will be mounted in 2004 in the new JET divertor.
Techniques are being developed to quantify deposition throughout the torus, and to establish the areas of
erosion. The surfaces of about 30 new tiles in a poloidal section of the divertor and main chamber are being
mapped to an accuracy of ~2 micron in the laboratory, and a similar number are being coated with markers.
Re-measurement after exposure in JET will determine the erosion/deposition pattern and lead to a much
improved knowledge of erosion/deposition and fuel retention in JET. [1] H-G Esser et al, Proceedings of the
Carbon Workshop, Jülich, Germany, Sept 2003 [2] M Mayer et al, ibid This work has been conducted under
the European Fusion Development Agreement and is partly funded by EURATOM and the UK Engineering
and Physical Sciences Research Council
Corresponding Author:
COAD, JOSEPH PAUL
Paul.Coad@jet.uk
EURATOM / UKAEA Fusion Association, Culham Science Centre, Abingdon, OXON OX14 3DB, U.K.
170
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-428
VULNERABILITY OF OPTICAL FIBERS FOR PLASMA DIAGNOSTICS
OF LASER MEGAJOULE
S. GIRARD (1), B. BRICHARD (2) J. BAGGIO (1) J-L. BOURGADE (1) M. DECRÉTON (2) F. BERGHMANS (2)
(1) CEA/DIF, BP n 12, 91680 Bruyères-le-Châtel, France (2) SCK-CEN Belgian Nuclear Research Center, B-2400 Mol,
Belgium
The Laser Megajoule (LMJ) project is a major component of the French simulation program to study nuclear
fusion by inertial confinement. Optical plasma diagnostic systems of the LMJ have to resist to the harsh
environment of this facility. The LMJ is able to focalise up to 1.8 MJ ultraviolet laser light into a volume of
few mm3 and this high density will drive a large amount of X-rays and nuclear particles (neutron and gamma
rays) when DT filled glass microballon implosion experiments are performed. Two different effects limit the
fiber integration in the diagnostic systems: the radiation-induced attenuation (RIA) and radiation-induced
luminescence. After many years of research, radiation-resistant multimode optical fibers have been developed
for steady state radiative environment representative of other facilities (e.g. ITER). Optical fibers with pure
silica cores based on SSU, STU or KU silica glasses and fluorine-doped claddings, exhibit low RIA levels.
Furthermore, their responses could also be improved by appropriate pre-treatments, like hydrogen-loading or
pre-irradiation. These fibers seem to be interesting candidates for the LMJ plasma diagnostics. However,
previous studies established that pure silica core fibers, with radiation-hardened properties under gamma-ray
irradiation, could present strong RIA levels for the short times after a transient exposure (duration of few
nanoseconds as in the case of LMJ). We will characterize the radiation behaviors of these “rad-hard” optical
fibers after a X-ray pulse (dose <1 kGy, dose rate > 1 MGy/s) and evaluate their vulnerability for the LMJ
facility. Comparisons with similar studies conducted for magnetic fusion diagnostic design for ITER will be
presented. This work was possible thanks the support of E.R. Hodgson from Euratom/CIEMAT
Corresponding Author:
S. GIRARD (1)
sylvain.girard@cea.fr
CEA/DIF, BP n 12, F91680 Bruyères-le-Châtel, France
171
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-433
APPLICATION OF ORTHOGONALLY POLARIZED TWOFREQUENCY LASER TO POLARIMETER FOR MAGNETIC FIELD
MEASUREMENTS OF LONG-PULSED FUSION DEVICES
TSUJI-IIO SHUNJI, MIYAZAKI TAKESHI, HAYAKAWA KAZUHIRO, AKIYAMA TSUYOSHI*, TSUTSUI
HIROAKI, SHIMADA RYUICHI
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550,
Japan *National Institute for Fusion Science, 322-6, Oroshi-cho, Toki-shi, Gifu 509-5292, Japan
We have developed a heterodyne magneto-optic polarimeter for magnetic fields measurements on long-pulsed
fusion devices to avoid the problems of zero-points drifts of long integration with commonly used pick-up
loops. When frequency shifters such as acousto-optic modulators are used to apply heterodyne techniques, the
alignment accuracy of recombined twoÅ@beams limits the precision of long-time field measurements. Hence
a transverse Zeeman laser, which emits two orthogonally polarized beams with slightly different frequencies,
was adopted as a light source. Since the two beams are perfectly collinear without any adjustments, the
accuracy of long-time measurements was significantly improved. However the instability of the laser
frequency induced by optical feedback from a fiber coupler to the laser source degrades the polarimetric
measurement when a polarization-maintaining fiber was used to freely propagate the orthogonally polarized
beams to sensor locations. To minimize the optical feedback, an optical isolator for the orthogonally polarized
two-frequency laser was tested. Since the key element of the optical isolator, a Faraday rotator, is sensitive to
temperature, the control of the rotator temperature was found to be crucial to keep good isolation. The
achieved accuracy of the Faraday rotation angle was less than 0.01 degrees, which corresponds to a resolution
of magnetic field measurement of 2 G in the case that a 40-mm long flint glass (OHARA PBH71) is used as a
sensor. Magnetic field measurement up to 5 T was tested using a superconducting magnet and nonlinearity of
measurement due to elliptic polarization of the beams will be discussed.
WITHDRAWN
Corresponding Author:
TSUJI-IIO SHUNJI
siio@nr.titech.ac.jp
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku,
Tokyo 152-8550, Japan
172
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-437
THE TEXTOR DIAGNOSTIC DATA MANAGEMENT CHAIN
KROM, J.G., KORTEN, M.K. KOSLOWSKI, H.R. KRAEMER-FLECKEN, A.
During the 2002-2003 shutdown of the Textor device many components of the Textor diagnostic data
management system were upgraded. Following the summer of 2003 Textor successfully resumed operation
with these new and upgraded components in place. This paper intents to give an overview of these components
and how they co-operate to form one integrated Data Management system. The components in question are: =
JDAQ: The Java (or Juelich) Data Acquisition System. A successful proof of this new Textor data acquisition
system has been achieved during the commissioning of Textor with the Dynamic Ergodic Divertor (DED).
This new system replaced the greater part of legacy RT2 systems used so far. Embarking from the well
established design principles of RT2, JDAQ aimed at an open, distributed and scalable system. It has been
completely written in the JAVA object oriented programming language supporting a homogeneous and strictly
modular software design, using native interfacing only to connect operating system specific hardware drivers.
JDAQ is designed as a multi-tier layered system, which can be run on a single node or distributed over a
network. = CSF & TPD: The TEC Common Storage Facility and the Textor Physics Database. A new data
storage facility was brought into operation, designed to contain the "raw" data obtained from JDAQ and other
acquisition systems. A separate part of this store contains the more more physics orientated data. = TWU: The
TEC Web-Umbrella All data stored in CSF, TPD and other Textor related databases is now accessible via one
common data access scheme. This "TEC Web-Umbrella" is based on the widely used HTTP protocol. =
Chain1: A chain of automatically run analysis software. The data access via the TWU scheme, allowed us to
build a framework of analysis programs that get executed whenever a particular data set has been collected by
JDAQ. Diagnosticians and/or plasma physicist, can provide their own analysis code, in their preferred
programming language, and plug this program into the Chain1 framework to provide data for the TPD. We
also intent to describe our experiences with this integrated system during this year of operation.
Corresponding Author:
KROM, J.G.
J.Krom@fz-juelich.de
Institute for Plasmaphysics, Forschungszentrum Juelich, D-52425, Jelich, Germany
173
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-446
NEW LOW LOSS TRIAXIAL AND MAGNETICS DIAGNOSTICS
FEEDTHROUGH AT JET
DIRKEN, PETER, LAM, NORMAN ALTMANN, HENK
Euratom-UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK
During 2003 the UKAEA designed a novel low loss tri-axial and magnetics feedthrough to be installed during
the 2004/2005 shutdown. The new low loss tri-axial and magnetics feedthrough incorporates two interacting
systems in one assembly. The two diagnostic systems supported by this feedthrough are: -Toroidal Alfvén
Eigenmodes [TAE] antennae (high frequency at high voltage and current). -Magnetic pick-up sensors
(milliVolts and milliAmps). Primary design considerations were: -Each TAE antenna will operate at up to
500kHz, 600V and 30A maximum. -RF interference from the TAE should not affect the signals from the
magnetic pick-up sensors to the extent that the signal-to-noise ration is significantly degraded, particularly
those used for MHD measurement. -For this purpose a new flexible tri-axial cable had to be designed,
prototyped, developed, manufactured and tested. The cable was required to operate reliably up to 450 degrees
C in UHV. -The design had to use materials compatible with the anticipated gamma and neutron radiation
levels in JET as well as providing a double tritium boundary. -All in-vessel sockets, plugs, conduits and
cabling have to be installed by Remote Handling installation to minimise the manual intervention time and
activities. This required remote operation of photogrammetry, welding, alignment and positioning systems and
processes. The paper will cover the constraints, design, manufacture, assembly and Remote Handling mock-up
trials of the feedthrough in preparation for in-vessel installation, ready for the intended use of the JET facilities
beyond 2004. Enhancements take into account the likely International Thermonuclear Experimental Reactor
(ITER) operating scenarios, which will be a considerably larger tokamak device. Additionally, design choices
are being made for ITER-relevant subsystems, which could be finalised even after the start of construction of
ITER. Diagnostic systems for ITER will necessarily be designed to take into account the requirement to
minimise manned intervention by installing systems by Remote Handling.
Corresponding Author:
DIRKEN, PETER
pdirken@jet.uk
Euratom-UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK
174
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-451
THERMO-STRESS ANALYSIS OF OPTICAL MATERIALS FOR HIGH
HEAT FLUX APPLICATIONS
OGORODNIKOVA OLGA, R. KÖNIG2), J. LINKE1),.G. PINTSUK1)
1)IWV-2, FZ-Juelich GmbH, EURATOM ASSOCIATION, D-52425 Juelich, Germany 2)Max-Planck-Institut fuer
Plasmaphysik, Greifswald, EURATOM ASSOCIATION, D-17491 Greifswald, Germany
The ideal optical material for high heat flux applications would have negligible absorption, scattering and
birefringence throughout the 0.15-5 micron band even at elevated temperatures. A low index of refraction is
also desirable to minimize transmission losses from surface reflections. The materials should maintain
acceptably low level of radiation emission and low thermally induced stresses to prevent failure by power load.
So, the ideal material should have high strength, high thermal conductivity, low elastic modulus and low
thermal expansion even at elevated temperatures. All of these optical and mechanical characteristics are
desirable for a reasonable price. In the present work, the optical materials investigated for
ultraviolet/visible/infrared window under high heat load are Al2O3, SiO2, ZnSe, ZnS, CaF2, MgF2, BaF2 and
CVD-diamond. All of them have a high transmission range. Finite element method has been used for
calculations of temperature and stress distributions. Influence of (1) power load, (2) pulse duration, (3) cooling
conditions, (4) atmospheric pressure on unloaded side, (5) surface and bulk heating and (6) size of the window
has been studied in detail by numerical calculations. The result of this study allows to choose the most suitable
window for different diagnostics in the next step fusion devices such as the stellarator W7-X. This procedure is
also qualified for the selection of windows materials for aerospace applications and high power continuouswave CW lasers. The investigations show in which operating regimes with regard to pulse duration, power
load and sizes every material can be used until it fails. For example, actively cooled sapphire window is
suitable for large window diameters (about 13 cm) which could cope with the expected maximum radiation
power loads for W7-X of 50 kW/m2 for more than 20 minutes. Other materials like fused silica, MgF2, ZnS
and ZnSe can only be used for smaller diameter windows (less than 5 cm). CaF2 is unacceptable for the
protective window for W7-X because of strong in-plane distortions during the long heat load. CaF2 can be
used only at low power load.
Corresponding Author:
OGORODNIKOVA OLGA
o.ogorodnikova@fz-juelich.de
IWV-2, Forschungzentrum Juelich, 52425 Juelich, Germany
175
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-458
DESIGN AND MANUFACTURE OF THE UPPER COILS AND OUTER
POLOIDAL COILS SUBSYSTEMS FOR THE JET-EP MAGNETIC
DIAGNOSTIC
PERUZZO SIMONE, W.BAKER(1), V.COCCORESE(2), T.EDLINGTON(3), S.GERASIMOV(3), S.HUNTLEY(3),
N.LAM(3), A.LOVING(3), N.POMARO(1), AND JET-EFDA CONTRIBUTORS
(1) Consorzio RFX - Association EURATOM-ENEA, C.so Stati Uniti 4, I-35127 Padova. (2) Consorzio CREATE Association Euratom-ENEA, Via Claudio 21, I-80125 Napoli. (3) UKAEA/Euratom Fusion Association, Culham Science
Centre, Abingdon, OX14 3DB, UK
The enhancement project of the magnetic diagnostics aims at the design, procurement, installation and
commissioning of new sets of magnetic transducers to be installed inside and outside the vessel, in order to
substantially improve the current capabilities of the JET magnetics. All in-vessel coils have been designed
considering the necessity of installing them by means of the Remote Handling system available at JET. The invessel sensors are grouped in 3 different sub-systems of two field component pick-up coils, to be located as
near as possible to the plasma, assembled on rails in order to ease Remote Handling installation. The 3 subsystems (Upper Coils, Outer Poloidal Limiter Coils, Divertor Coils) are attached to different structures of the
first wall and replicated for redundancy in 2 Octants, for a total of 76 new pick-up coils. This paper will focus
on the detailed design and manufacture of the first two subsystems (UC and OPLC). The final design of the
two sub-systems has been carried out in order to fulfil all the operation and fault condition design criteria for
JET in-vessel components. Preliminary tests on mock-ups have been performed in order to guarantee the
compatibility with the features of the Remote Handling system. Most of the coils are made of a Mineral
Insulated Cable (MIC) wound around Inconel formers, which give a good reliability in the JET environment.
The relatively low frequency response (about 50 kHz) is more than adequate for plasma control and
equilibrium reconstruction, which require a maximum frequency of 10 kHz. In addition a sub-set of 14
“tangential coils”, located on the Outer Poloidal Limiter, is designed as titanium bare wire wound onto an
alumina ceramic former, so that they can be used for high frequency applications (e.g. MHD studies). For low
frequency coils special design attention has been dedicated to the coils terminations, connecting the MIC to
standard shielded cables inserted into conduits. The signals are finally driven out of the vessel by means of
new feedthroughs, which have been designed in order to minimize interference with other systems. To
improve reliability all coils are subjected to severe vacuum and temperature tests. In addition a calibration
procedure is applied to all coils, in order to minimise systematic measurement errors.
Corresponding Author:
PERUZZO SIMONE
simone.peruzzo@igi.cnr.it
Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova Italy.
176
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-459
DESIGN OF EX-VESSEL MAGNETIC PROBES FOR JET-EP
CHITARIN GIUSEPPE, F. BASSO (1), V. COCCORESE (2), S. PERUZZO (1), N. POMARO (1), T.
EDLINGTON(3), C.SOWDEN (3), S.CRAMP (3), K.FULLARD (3) AND JET-EFDA CONTRIBUTORS
(1) Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, Padova, Italy (2) Consorzio CREATE, Association
Euratom-ENEA, via Claudio 21, Napoli, Italy (3) UKAEA-Euratom Fusion Association, Culham Science Centre, Abingdon,
OX14 3DB, UK
The enhancement of JET magnetic diagnostics will include the installation of new ex-vessel sensors, which are
mainly intended to provide information useful for the characterization of the iron transformer during the
discharge. The numerical codes used for equilibrium reconstruction need an equivalent, axisymmetric iron
model and inaccuracies in the iron core modelling may affect the accuracy of the equilibrium reconstruction,
especially during some critical phases of the discharge. The present numerical models of the iron core are
based on the field measurements obtained by time-integrated pick-up coil and flux-loop signals, together with
simplified information on the geometry and magnetic properties of the iron structure. However, no information
is presently available on the magnetic field configuration in proximity of the iron core structure and on the
contribution of the residual magnetization of the iron. It is expected that the equilibrium reconstruction should
benefit from the absolute measurements of the iron stray field before and during the pulse. The ex-vessel
sensor system to be installed consists of a number of Hall sensors, complemented by “local” pick-up coils and
“octant average” flux loops. All these sensors are supported by rails to be attached to the iron core structure.
There are in total 26 new sensors grouped in 2 subsystems, named Collar probes and Limb probes,
respectively. The design activity is conducted in collaboration between the Association EURATOM-ENEA
and the JET Operator. For these purposes new Hall probes are being introduced in the magnetic diagnostics.
Some prototype Hall probes have already been installed on the Limb surface and tested before the start of the
2004 shutdown, mainly in order to verify the reliability of the absolute field measurements over a time-scale of
several weeks and to compare the behaviour of different kinds of Hall sensors under real operating conditions.
The paper will describe the preliminary results obtained with Hall probe prototypes, and will discuss the
choice of type and location of Collar and Limb probes, together with other design and manufacturing issues.
Corresponding Author:
CHITARIN GIUSEPPE
giuseppe.chitarin@igi.cnr.it
Consorzio RFX, C.so Stati Uniti 4, 35127 Padova, Italy
177
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-466
TRANSDUCERS AND SIGNAL CONDITIONERS OF THE RFX NEW
MAGNETIC MEASUREMENT SYSTEM
POMARO NICOLA, BASSO FRANCESCO
Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova, Italy.
The paper presents the out of vessel magnetic measurement system recently installed on RFX. The system has
been completely redesigned in order to meet the electromagnetic, mechanical and thermal requirements of the
new toroidal assembly of RFX, with the thinner shell and the new system of coils for MHD modes control.
The system allows to measure the main integral plasma quantities: toroidal current, loop-voltage, toroidal flux.
Special partial poloidal voltage measurements are foreseen to allow the study of halo currents poloidal and
toroidal spectra. Furthermore, 224 two axis probes and 240 saddle loops were installed for local field
measurement. In total, 744 independent measurements are available; 712 of them are placed in between the
vacuum vessel and the new thin shell, where an air gap only 6 mm thick is available, temperature can rise up to
200 C, and a voltage up to 2000 volt can develop between shell and vessel during machine operation. To
comply with such extreme requirements, specific techniques were developed for probes and connections,
which make use of advanced polymers for protection and insulation. Particularly challenging was the design
and realisation of Rogowski probes, which were placed into existing grooves in the vessel, only 3 mm thick.
To avoid mechanical interferences, a detailed study of cable paths was carried out, with the aid of 3D
computer modelling. To obtain the best precision in probes and connections installation, a special tool was
designed and realised, which allowed to draw directly on the vessel surface the position of each probe and
cable, with a precision better than 1 mm. Conditioning electronics was completely redesigned to improve
accuracy and interferences immunity. Integrators drift was lowered by an order of magnitude with respect to
previous system, and a new Front-End was designed to cope with extended signals bandwidth and the presence
of high frequency noise due to new powerful switching power supplies. Also mechanical characteristics of
conditioning channels were changed to improve modularity and reliability. This paper describes the realisation
of all integral probes, the tools and procedures adopted for installation and tests of the whole probe system,
and the design and realisation of the conditioning electronics.
Corresponding Author:
POMARO NICOLA
nicola.pomaro@igi.cnr.it
Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova, Italy
178
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-476
WIDE-ANGLE INFRARED THERMOGRAPHY FOR JET-EP
ERIC GAUTHIER, E. THOMAS, B. BERTRAND, P. CHAPPUIS, L. DOCEUL, D. GUILHEM, M. MISSIRLIAN, P.
ANDREW2, P. COAD2, T. TISCORNIA2, C. ANTONNUCCI3, C. DAMIANI3, J. GAFERT3, A. LIOURE3 AND
CONTRIBUTORS TO THE EFDA-JET WORK PROGRAM
2Euratom/UKAEA Fusion association, Culham Science centre, Abingdon OX14 3DB, UK 3EFDA-CSU Culham, Culham
Science centre, Abingdon OX14 3DB, UK
The surface temperature of the plasma facing components need to be measured to operate the tokamak in a
safe manner and to calculate the power flux impinging on the different parts of the machine. In the frame of
JET-EP (Enhancement Phase), a new infrared thermography diagnostic is being developed. The objective is to
provide a wide-angle view in the infrared range (3 to 5 µm) for thermography in the main chamber and
divertor aiming at real time machine protection and for analysis of the power flux deposition during normal
operation and transient events such as disruptions and ELMs. The diagnostic will be able to measure
temperature with a large dynamic range from operating temperature of 200 C up to a maximum temperature of
2000 C. The enhanced dynamic range is achieved by using a multi-exposure time: acquisition with three
different exposure times is performed and the corresponding frames are combined in a single thermal image. In
order to measure accurately the power and energy deposition during ELMs, a time resolution of the order of
100µs is achieved by reducing the image size to 128x8 pixels, and by using a 40 MHz pixel clock. In order to
image a large section of the tokamak in both poloidal and toroidal directions, dedicated optics have been
designed. The optics have a field of view of 70 degrees, viewing the divertor, the inner wall, the outer poloidal
limiters, the ITER-like ICRH antenna and the top limiter. The main feature of the optics design is to be ITERrelevant. To this end, the optical components are based mainly on reflective optics: the only kind which can
sustain high neutron radiation. The optical system consists of an endoscope installed in a lower limiter guide
tube, a Cassegrain telescope and a relay group of lenses, the latter being connected to the camera body. The
design uses a concave aspheric mirror located behind a plan mirror equipped with a small aperture. Both
mirrors are made of stainless steel coated with gold. The Cassegrain telescope is composed by elliptic and
hyperbolic mirrors, both made of Zerodur® glass (gold coated). Finally, the image is magnified and
transmitted to the detector with 4 Silicon and Germanium relay lenses. Strong requirements in positioning the
optical elements, in addition to the vacuum and thermal constraints, imposed a complex and challenging
mechanical design. The paper will present the main characteristics of the optical and mechanical designs.
Corresponding Author:
ERIC GAUTHIER
e.gauthier@cea.fr
Association EURATOM-CEA, CEA/DSM/DRFC, CEA Cadarache, 13108 Saint Paul Lez Durance (France)
179
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-477
LITHIUM BEAM DEVELOPMENTS FOR HIGH-ENERGY PLASMA
DIAGNOSTICS
ANDA GABOR, S. BATÓ G. PETRAVIC S. ZOLETNIK
Lithium beam developments for high-energy plasma diagnostics G. Anda, S. Bató, G. Petravic, S. Zoletnik
KFKI-RMKI, Association EURATOM, P. O. Box 49, H-1525 Budapest, Hungary The injection of 10-100
keV Li neutral beam into magnetically confined fusion plasmas causes collisionally induced Li line emission
at 670.8 nm. Observing the intensity and the fluctuations of this Li resonance line along the beam, it is possible
to reconstruct the density profile and the 2 dimensional correlation of the electron density fluctuation.
Measurement of the line shape gives information on the magnetic field in the observation volume. Charge
exchange with plasma impurity ions gives the possibility of the determination of impurity ion concentrations.
In this paper the JET Li-gun and a new ion optic geometry are investigated experimentally and with the CPO
simulation code. The beam has been tested in the laboratory at IPP-Garching to find optimal operation
conditions and limits. Intersecting the beam path with a Titanium plate the beam ions get neutralised on the
surface and some of them become excited. The beam profile was measured by observing the radiation resulting
from this excitation. Although the photon yield from one ion (or atom) is unknown, it can be assumed that the
individual ions radiate independently and the observed light profile is proportional to the beam current profile.
These test results have been used to validate the code calculations and to explore possible beam source
upgrades. A new ion optic geometry is developed to determine the relation between the emitter temperature
and the drawn current and to test a new type of emitter and emitter material. It turned out that the drawn
current highly depends on the temperature of the emitter. The recently developed new emitter material was
found to be capable of delivering substantially higher ion current than the conventionally used B-eucriptit and
spodumen sources. These developments promise much better signal to noise ratios and potentially new
application areas for high-energy Lithium beam diagnostics.
Corresponding Author:
ANDA GABOR
andag@rmki.kfki.hu
H-1121 Budapest Konkoly Thege Miklos Street 29-33.
180
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-502
THE NEW TAE - ALFVÉN WAVE ACTIVE EXCITATION SYSTEM AT
JET
DUCCIO TESTA, A.FASOLI1,2, P.BEAUMONT3, R.BERTIZZOLO1, M.BIGI3, C.BOSWELL2, R.CHAVAN1,
S.HUNTLEY3, N.LAM3, A.LOVING3, S.MILLS3, V.RICCARDO3, S.G.SANDERS3, J.A.SNIPES2, J.THOMAS3,
P.TITUS2, L.VILLARD1, M.VINCENT3, R.WALTON3, M.WAY3, AND JET-EFDA CONTRIBUTORS
(1) CRPP, Association EURATOM-Confédération Suisse, EPFL, 1015 Lausanne, Switzerland (2) Plasma Science and Fusion
Center, MIT, Cambridge, MA 02139, USA (3) EURATOM-UKAEA Fusion Association, Culham Science Centre, OX14 3EA,
Abingdon
After many years of successful operation, the JET saddle coil system will be dismantled during the next
shutdown. A new antenna system has been designed to replace it and excite MHD modes in the Alfvén
frequency range (10-500kHz), keeping similar operational capabilities (I~30A, V~1kV, P~5kW). Due to their
geometry, the saddle coils could drive only low toroidal mode numbers, n=0-2. Conversely, the n’s that can be
driven unstable in ITER by fusion generated alphas or other fast particles are expected to be in the range n~520. The mismatch between the modes driven by the saddle coils and those that are made unstable by the fast
particles is already observed on JET, which sometimes makes it difficult to extrapolate the present results to
large burning plasma experiments. The new antenna system is designed to overcome this limitation. It
comprises two assemblies of four toroidally spaced coils each, situated at opposite toroidal locations. Each coil
is made using 18 turns of 4mm Inconel 718 wire, covers a toroidal and poloidal extent of ~25cm, and is
individually insulated from the supporting frame with Shapal-M spacers. The first turn sits approximately
40mm behind the poloidal limiters. The coils are mounted on a 3mm-thick Inconel 625 open structure, to avoid
a closed path for disruption-induced currents. This structure is attached to the poloidal limiters and the remains
of the saddle coil brackets with four attachment points, so as to optimise the load distribution, and it is further
protected by CFC tiles. Any combination of 4 out of the 8 antennas can be chosen to excite different n-spectra,
up to n~20. Code calculations show that this arrangement provides a coupling to the plasma for a n=5 TAE
that is of the same order as that achieved with the present, much bigger, saddle coils for an n=2 TAE for the
same JET equilibrium. Thus, it is foreseen that the real-time tracking capabilities of the old saddle coil system
will be maintained and, possibly, control of the marginal stability limit could be envisaged for the intermediate
n-modes driven by the new antennas. In addition to the constraints imposed by halo current and disruptioninduced voltages and currents, the design must comply with the requirements of a remote handling installation.
Design principles and constraints will be presented along with the results of the coupling and engineering
analysis, and a discussion of the possible extrapolation of such a system to ITER
Corresponding Author:
DUCCIO TESTA
dtesta@jet.uk
CRPP, Association EURATOM-Confédération Suisse, EPFL, 1015 Lausanne, Switzerland
181
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-503
NEW MILLIMETER-WAVE ACCESS FOR JET REFLECTOMETRY
AND ECE
CUPIDO LUIS, L. CUPIDO(1), E. DE LA LUNA(2), ET AL FROM CSU(3), FOM(4), CNR(5), IPP(6), CFNIST(1), JET-EFDA(7)
1-CFN, IST, Lisboa, Portugal/ 2-CIEMAT, Madrid, Spain/ 3-CSU Culham Science Centre, Abingdon UK/ 4-FOMRijnhuizen, The Netherlands/ 5-IFP-CNR Milano Italy/ 6-IPP, Max-Planck-Institut, Garching, Germany/ 7-UKAEA
Abingdon UK.
Millimeter-wave ECE and Reflectometry at JET employ state of the art electronics but are limited in
performance by the existing waveguide and antenna system. The use of long runs of waveguides with high
losses and non optimized antennas lead to difficult measurement conditions for both ECE and reflectometry.
An access system has been designed to improve the performance of reflectometry and enable the installation of
antennas for ECE oblique view. The new antennas will allow the ECE radiation to be collected at different
angles with respect to the magnetic field, which is extremely useful to improve the interpretation of ECE
temperature measurements in JET. For reflectometry there is an urgent need to improve the edge density
measurements as both the lithium beam and Thomson scattering exhibit limitations of resolution at lower
densities. The project aims at the installation of a millimeter-wave access system consisting of six
antennas/waveguides for probing the mid-plane of the JET plasma, covering a frequency range of 60-190GHz.
Two of those antennas and transmission lines are dedicated to the oblique view ECE. Access to the plasma will
be done using a port with direct line of sight to the plasma, a limiter guide tube. This port allows a complete
bundle of antennas and waveguides to be inserted from the outside of the vessel. The paper presents the
analysis and design of the antennas, corrugated waveguides, vacuum windows and instrument interface. Four
antenna apertures take advantage of the excellent coupling of the propagating HE11 waveguide mode to the
free-space Gaussian beam which is also inherently broadband. A vacuum boundary with double dielectric
windows and inter-stage vacuum was designed to operate in the 60-190GHz range. The corrugated waveguides
were designed for the 60-190GHz range. Oblique ECE measurements, with a wider frequency range, 100400GHz, use smooth circular waveguides. The coupling to the fundamental waveguides of the instruments is
provided by quasi-optical coupling to fundamental waveguides using matching mirrors and horn antennas.
Quasi-optical units are built in such a way to allow for flexible configuration of input/output ports along with
the capability of adjusting polarization and frequency separation. The overall setup results in an improvement
of about 30dB for reflectometry that will enable broad band reflectometry for density profile measurement and
ECE oblique experiments to be performed for the first time on JET.
Corresponding Author:
CUPIDO LUIS
cupido@mail.ua.pt
IST, Centro de Fusão Nuclear, Instituto Superior Técnico, 1049-001 Lisboa, Portugal
182
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-504
CONTROL PROCESS STRUCTURE OF ASDEX UPGRADE´S NEW
CONTROL AND DATA ACQUISITION SYSTEM
RAUPP, GERHARD, G. NEU, W. TREUTTERER, V. MERTENS, D. ZASCHE, TH. ZEHETBAUER
Selected also for oral presentation
O2B-D-504
ASDEX Upgrade´s new real-time CODAC was designed to demonstrate state-of-the art plasma control and
operation methods required for advanced fusion machines. It is an open distributed system of controller and
data acquisition nodes exchanging process data via a common real-time network. The software has two layers
with underlying universal infrastructure functions and superior task specific application processes. The
infrastructure provides signal exchange methods, alarm and log mechanisms (required on each node), and
specific central processes for time and cycle management, system self- monitoring, and protocol extraction.
The application layer consists of various application processes freely allocated onto controller nodes. The
initial control implementation breaks down into: - plasma FF&FB with two processes, one for the feedback of
position and shape via PF coils, and one for performance control with gas and heating systems - monitoring of
machine and plasma with specific processes to check plasma position & shape, plasma performance &
instability, machine protection systems, power supplies, coil currents, coil stress, and a supervisor process to
take top-level decisions about the discharge sequence - data generation with processes to exchange signals
with actuators or sensors, to compute equilibria and other , to evaluate complex input data, and to generate
real-time reference values We will present the structure of the control processes, and show how these
cooperate to form feedback control loops, to monitor machine and plasma, to take intelligent decisions about
the discharge evolution, and to interact with protection systems to terminate the discharge in case of normal
and abnormal operation.
Corresponding Author:
RAUPP, GERHARD
raupp@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, D-85748 Garching,
Germany
183
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-506
MULTI-SUPPORT VECTOR MACHINES FOR DISRUPTION
CLASSIFICATION IN TOKAMAK REACTORS
MARIO VERSACI, ANTONINO GRECO FRANCESCO CARLO MORABITO
Facoltà Ingegneria Università Reggio Calabria Via Graziella Feo di Vito - I-89100 Reggio Calabria Italy
Disruption is a sudden loss of magnetic confinement that can cause damage to the machine walls and support
structures. For this reason, it is of practical interest to be able to detect the onset of such an event early. The
prediction problem can be expressed in terms of classification. Particularly, when a shot starts, it is imperative
to know the kind of disruption. In this paper, a novel technique of classification of plasma disruption in
tokamak reactor (JET) which use Support and Multi-Support Vector Machines (SVMs, M-SVMs) with Multi
Layer Perceptron Neural Networks (MLPNNs) and Learning Vector Quantization (LVQ) is presented.
Actually, in scientific literature, there are two methods for Multi-class Support Vector Machines. The first one
is made by combination of different binary classifiers. The second one considers just one optimised relation.
M-SVMs can be considered as a natural extension of SVMs for multi-class classification; in fact, SVMs
classify by means a binary approach (two classes only) while M-SVMs solve problems in which data can
belong to several classes. Training and testing data sets have been made choosing time sampling and particular
kinds of shots for classifying disruptions. In addition, to improve the quality of classification, special signals
have been take into account for our purpose (plasma current, mode lock,…). The obtained results show the
goodness of the proposed approach, with respect to MLPNNs, in terms of percentage of false and missing
allarms. Moreover, the reduced computational complexity of M-SVMs is useful for on-line applications
especially in the case in which the classification of shot is a step of control task.
Corresponding Author:
MARIO VERSACI
versaci@ing.unirc.it
Facoltà Ingegneria Università Reggio Calabria Via Graziella Feo di Vito - I-89100 Reggio Calabria Italy
184
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-513
OPTICAL DESIGN OF THE OBLIQUE ECE ANTENNA SYSTEM FOR
JET
CARLO SOZZI (1), ALESSANDRO BRUSCHI (1) ALESSANDRO SIMONETTO (1) ELENA DELALUNA (2) JOHN
FESSEY (3) VALERIA RICCARDO (3) AND JET-EFDA CONTRIBUTORS
(1)Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, Milano, Italy (2)Asociación EURATOM-CIEMAT,
CIEMAT, Madrid, Spain (3)EURATOM-UKAEA Association, Culham Science Centre, Abingdon, UK
The correct measurement of the plasma temperature has been an important issue since the beginning of
thermonuclear fusion research. The introduction of Electron Cyclotron Emission (ECE) diagnostics as
routinely available electron temperature measurement in tokamak and stellarators provided a major burst in the
understanding of the plasma behavior in fusion relevant conditions. The availability of effective and powerful
additional heating systems opened very high temperature scenarios bringing to light new phenomena related to
the electron temperature measurements, actually arising the question of what exactly the diagnostic itself is
measuring. In particular some systematic disagreements between ECE and Thomson Scattering diagnostics
have been observed in the presence of NBI and ICRF heating (TFTR and JET), and ECR heating (FTU),
probably due to deviation of the electron population from Maxwellian-bulk distribution. These observations
have substantiated the proposal of the so called Oblique ECE diagnostics on JET, in which the ECE radiation
is detected along lines of sight outside the poloidal plane. This layout allows the study of the electron
distribution function at low energies revealing any non Maxwellian shape. This paper is devoted to the design
of the quasi optical antenna for the Oblique ECE diagnostic. The physics requirements imply two lines of sight
at about 10 and 20 degrees respectively in the toroidal direction. Severe geometrical constraints are imposed
by the mounting method of the antenna, inserted in the vacuum vessel together with the group of six oversized
waveguides devoted to Reflectometry and Oblique ECE itself and their surrounding structure. The two
Oblique ECE waveguides share the same horizontal plane with one reflectometer waveguide, and are at the
opposite side of the structure with respect to this. The antenna will be built with three flat mirrors and an
ellipsoidal one, the last being shared by the two lines of sight. The mirror arrangement was optimized using
electromagnetic calculations performed at several frequencies in the range of work foreseen for the diagnostic,
extending from 100 to 400 GHz.
Corresponding Author:
CARLO SOZZI (1)
sozzi@ifp.cnr.it
Istituto di Fisica del Plasma del CNR - Via R.Cozzi, 53 - 20125 Milano - ITALIA
185
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-515
ITER DIAGNOSTICS: MAINTENANCE AND COMMISSIONING IN THE
HOT CELL TEST BED
WALKER CHRISTOPHER I., A.E.COSTLEY, R.GOTTFRIED(1), B.HAIST(2), K.ITAMI, T.KONDOH,
G.D.LOESSER(3), J.PALMER(4), T.SUGIE, A.TESINI, G.VAYAKIS
ITER International Team, (1) Framatome, (2) Oxford Technologies, (3) PPPL, (4)EFDA
ITER diagnostic equipment is integrated in 6 equatorial and 12 upper ports, 16 divertor cassettes and 5 lower
ports. Diagnostic equipment in these locations is designed to be removed and then repaired, tested and
commissioned in the hot cell area. In this paper the requirements and methods of repair and testing on these
components are described. Design features that facilitate repair are included in diagnostic port plugs etc.
Appropriate reception testing allows a repair strategy to be formulated to minimize hot cell time. All
equipment to be reinstalled is checked as acceptable before embarking on the complex remote handling
transport and installation procedure. At the hot cell a dummy port is provided for tests of mechanical and
vacuum integrity as well as commissioning of the diagnostic equipment. The scope of the hot cell maintenance
and commissioning activities is summarised and an overview of the integration of the diagnostic equipment is
given.
Corresponding Author:
WALKER CHRISTOPHER I.
walkerc@itereu.de
ITER International Team, Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2 , 85748
GARCHING,Germany
186
- D - Diagnostics, Data Acquisition and Remote Participation.
P2C-D-519
NEW BOLOMETRY CAMERAS FOR THE JET ENHANCED
PERFORMANCE PHASE
MCCORMICK KENT, A. HUBER(1) C. FUCHS(2) C. INGESSON(3) J. FINK(2) W. ZEIDNER(2) A. GUIGON(4)
S. SANDERS(1)
(1) EURATOM/UKAEA, Culham, Abingdon, UK (2) Max-Planck-Institut für Plasmaphysik, Garching, Germany (3) FOM
Institute for Plasma Physics, Nieuwegein, The Netherlands (4) Close Support Unit - EURATOM, Culham, Abingdon, UK
JET is an experimental fusion device, the largest in the world, now undergoing a major upgrade. This enables
replacement of the bolometer cameras – vertical and horizontal - used to register the temporal evolution and
spatial distribution of radiation emanating from the plasma. The need is based on inadequate spatial
coverage/resolution over the cross section. Namely, the plasma configuration has progressed from a state
where radiation was distributed around the circumference or at the inside wall (limiter plasma) to that where it
is concentrated within a poloidal region near the bottom of the vacuum vessel (diverted plasma). The current
vertical camera, in operation since 1984, has become increasingly inadequate for the task. In particular, it is
often impossible to produce satisfactory tomographic reconstructions of the divertor radiation pattern, although
of primary interest for divertor scenarios. The new vertical camera has 24 channels, 16 covering the entire
cross section and another set of 8 probing the divertor region - in contrast to 9-10 working channels of the old
vertical camera (initially 14 channels). The new horizontal camera has 24 channels (vs. 20 for the old),
including a subset of 8 for the divertor region and another 4 for the upper boundary. In addition to the
increased number of viewing cones optimized for contemporary JET plasmas, new bolometer detectors and
electronics are being employed, similar to those on ASDEX-Up, Tore-Supra and RFX. The new(old) detector
is an 8(4)m gold absorbing layer placed on a 20(7.5)m mica(kapton) foil with a gold resistance meander on the
backside. The thicker layer effects detection of higher energy quanta over 5eV-8keV. The 1.2kohm meander
permits a higher operation voltage (40Vp-p vs. 10V). This, together with use of phase-locked techniques at a
carrier frequency of 50kHz (dc for old cameras) and an improved grounding/shielding concept, will lead to an
improved (>>10) signal-to-noise ratio and time resolution (<20ms). The paper will discuss salient aspects of
the new cameras - to be mounted on JET in Sept./Oct. - showing examples of the enhanced tomographic
capabilities. Design criteria demanded of in-vessel diagnostics on JET will be addressed: a) the ability to
withstand large forces associated with abrupt termination of the plasma current and b) the nearly reactor-level
quality control during construction - dictated by the tritium-phase of JET operation.
Corresponding Author:
MCCORMICK KENT
gkm@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching, Germany
187
- E - MAGNETS AND POWER SUPPLIES.
P2T-E-9
THE BATCH PRODUCTION FOR SUPERCONDUCTING MAGNET
COILS OF EAST (HT-7U)
GAO DAMING, CHEN SIYUE, YU JIE, WU JIEFENG, PAN YINNIAN, WU WEIYUE, LI BAOZENG, WEN JUN,
ZHANG PING, ZHU WENHUA, TAO YUMING, PAN WANJIANG, WU YU
Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126 Hefei 230031, P.R.China
Abstract: The construction of Experimental Advanced Superconducting Tokamak ( EAST ), approved by
Chinese government as one of Chinese national mega project of science research is smoothly underway
according to its schedule. The core of EAST Project is superconducting magnet coils (SMC), composed of 16
toroidal field coils (TFCs) and 14 poloidal field coils (PFCs). The prototypes for SMC had successfully been
completed and passed the cryogenic tests early of 2003. Since then the batch production for SMC have begun
at ASIPP. Up to now, 44 CIC conductors have been completed among the 58 CIC conductors in total; 26 half
TFCs and 8 PFCs have been completed separately among the 34 half TFCs and 16 PFCs including the model
and spare coils; other correlative procedures are also made at same time. Nearly 2/3 workloads for SMC have
been performed, and whole will be finished late of 2004. This paper emphasizing on the various technology
issues that must be faced and solved for 4 R&D lines of SMC after translating to batch production. To describe
the optimization of welding program for conduit joint, decreasing protrusion height inside, limiting insertion
gap, setting real NDT methods on CICC line. Also to describe construction of 2 new addition winding lines,
fabrication of U-shape current joint between 2 half TFCs, welding helium stub on conductor in site, protecting
cable over-heat and damage; sealing TF case, making twice VPI (vacuum pressure impregnation) treatments
for each TFC in special oven and once VPI treatment in site for each of four big diameter PFCs based on a
series of experiments; Quality control methods during batch production process for SMC are also discussed.
Corresponding Author:
GAO DAMING
dmgao@ipp.ac.cn
Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126 Hefei 230031, P.R.China
188
- E - Magnets and Power Supplies.
P2T-E-20
STUDY ON HIGH-POWER HIGH-FREQUENCY INVERTER FOR FAST
PLASMA POSITION CONTROL IN EAST SUPER-CONDUCTING
TOKAMAK
LIU ZHENG-ZHI, Y. YU(1), X. ZHANG(2), R. C. CHENG(1), S. S. LU(1), Z. WEI(1), C. W. ZHANG(2)
(1) Institute of Plasma Physics, Chinese Academy of Sciences, P. O. Box 1126, Hefei 230031, P. R. China (2) Hefei
University of Technology, Hefei 230009, P. R. China
In Experimental Advanced Super-conducting Tokamak (EAST*), the fast plasma position control is of
fundamental importance for suppression of inherent vertical instability of elongated plasma. An active control
system must be incorporated to compensate for resistive decay of the eddy currents of Passive Stabilizing Plate
(PSP) and to maintain plasma at a reference vertical position. The Power Supply for Fast plasma position
Control (FCPS) is to energize the active control system and to realize the fast current tracking under the
command from central control for plasma vertical position control in real time. The technical requirement of
FCPS and the parameters of the Internal Vertical Coils (IVC) in preliminary design will be introduced. The
preliminary design, digital simulation and principle experiment of FCPS will be presented. R & D on
AC/DC/AC converter topology and its control has been made progress. The current-source, three-phase PWM
rectifier (AC/DC) with current vector control, and the multi-parallel and phase-shifting PWM H-bridge
inverter (DC/AC) topology have been studied and developed. The principle experiment has been carried out
and it shows good agreement with the design and simulation. It has been shown that the high-power highfrequency current-source PWM converter (AC/DC/AC) will be satisfied with all the requirements of FCPS and
has some unique advantages as the followings: 1. Four-quadrant operation (Bi-directional current) 2. Fast
current tracking----up to 100HZ and even higher 3. Accurate regulation----Error of current tracking less than
5% 4. Easy for capacity enlargement and redundancy design----up to 10 MW 5. Multi-parallel and phaseshifting PWM to realize high frequency modulation with low switching frequency of devices 6. Sine wave and
Unit PF in AC input 7. High feasibility, high reliability and high flexibility *The former name of EAST was
HT-7U super-conducting Tokamak in CAS-IPP
Corresponding Author:
LIU ZHENG-ZHI
zzliu@ipp.ac.cn
P. O. Box 1126, Hefei 230031, P. R. China
189
- E - Magnets and Power Supplies.
P2T-E-23
A LOW COST JOINT FOR THE ITER PF COILS, DESIGN AND TEST
RESULTS.
STEPANOV BORIS, BRUZZONE PIERLUIGI, VOGEL MARTIN
CRPP, Villigen-PSI, Switzerland
The Poloidal Field (PF) coils of the International Thermonuclear Experimental Reactor (ITER) are designed
with use of NbTi cable-in-conduit conductors carrying a current up to 45 kA. Each PF coil consists of several
modules wound in double pancakes, joints connect that double pancakes electrically in series at the coil
periphery. In order to minimize the manufacturing cost maintaining a high reliability, a low cost joint design is
developed at CRPP and tested in SULTAN test facility in frame of ITER task for PF coils. The concept of the
joint is a single fully welded stainless steel box and one-pitch joint through sectioned copper saddle blocks.
The joint sample includes two full-scale joints, one in the hairpin configuration at the bottom of the sample and
one as an overlap joint at mid length of the sample. Lifting the sample in the SULTAN facility, either the
hairpin or the overlap joint can be placed in the center of the facility and tested under ITER-relevant operating
conditions. The low cost joint design for the ITER PF coils is described and the test results (DC and AC loss)
are presented in this paper.
Corresponding Author:
STEPANOV BORIS
boris.stepanov@psi.ch
CH-5232 Villigen PSI, WMHA/C23
190
- E - Magnets and Power Supplies.
P2T-E-30
130KV 130A HIGH VOLTAGE SWITCHING MODE POWER SUPPLY
FOR NEUTRAL INJECTORS - CONTROL ISSUES AND ALGORITHMS
GANUZA, DANIEL, GARCÍA, FRANCISCO (JEMA) ZULAIKA, MIKEL (JEMA) PEREZ, ALBERT (JEMA)
JONES, TIMOTHY (EURATOM/UKAEA FUSION ASSOCIATION)
JEMA: Paseo del Circuito 10. E-20160 Lasarte - Oria, Spain EURATOM/UKAEA Fusion Association: Culham Science
Centre, Abingdon, OX14 3DB UK
The company JEMA has delivered to the Joint European Torus (JET facility in Culham) two High Voltage
Switching Mode Power Supplies (HVSMPS), each rated at 130kVDC and 130A, which will feed the grids of
two PINI loads. This paper describes the main control issues and the algorithms developed for the project. The
most demanding requirements, from the point of view of the control are an absolute accuracy of +/- 1300V and
the possibility of performing up to 255 re-applications of the high voltage during a 20 second pulse. Keeping
the output voltage ripple to the specified tolerance has been a major achievement of the control system. Since
the output stage is formed of several modules (120) connected in series, their stray capacity to ground
significantly influences the individual contribution of each single module to the global output voltage. Two
complementary techniques have been used to balance the effects of the stray capacities. On the one hand, a
study has been carried out in order to find the optimum firing sequence of the 120 modules of the output stage,
considering the distribution of capacitances. On the other hand, an active ripple compensation algorithm has
been implemented. The fast re-applications requirement has a significant impact on the intermediate DC Link
stage. Such section is composed of a capacitance of 0.83 Farads at 650V, which feeds the 120 output stage
modules The DC Link is fed by a 12 pulse SCR rectifier, whose matching transformers are connected to the
36kV Grid. Every re-application and every voltage shutdown supposes a quasi-instantaneous power step from
zero to 17 MWatt load and vice-versa. Fast open loop algorithms have been implemented in order to keep the
DC Link voltage inside acceptable margins. Moreover, the HVSMPS output characteristics have to be
maintained during the rapid and important voltage fluctuations of the 36kV mains (28kV – 37kV). The general
control system is based on a Simatic S7 PLC, and a SCADA user interface. Up to 1000 signals are considered.
The control system has demonstrated to allow for a rapid and accurate identification of faults and their origin.
Corresponding Author:
GANUZA, DANIEL
d.ganuza@grupojema.com
JEMA. Paseo del Circuito 10. E-20160 Lasarte - Oria (Spain)
191
- E - Magnets and Power Supplies.
P2T-E-31
FIBERGLASS UNIDIRECTIONAL COMPOSITE TO BE USED FOR
ITER PRE-COMPRESSION RINGS
NARDI CLAUDIO, LIVIO BETTINALI (*) ALDO PIZZUTO (*)
(*) ENEA - Via E. Fermi 45 - 00044 Frascati (Roma)
The ITER magnet system will be kept in place by a system of pre-compression rings. These rings have
stringent requirements as far as the material requirements are concerned. They must assure the mechanical
strength and stiffness characteristics at both the cryogenic temperature (4¢XK) and room temperature. The best
solution, at present knowledge, is to make rings in composite material, using unidirectional glass fibers and a
resin matrix. ENEA performed a series of impregnations in order to study the characteristics of such a material
both at room temperature and at liquid nitrogen (77¢XK) temperature. The used glass was an S-2 glass,
because of his good mechanical characteristics, and the matrix was epoxy resin. The glass, in form of fibers
having 14 ƒÝm diameter and tex 725, have been wounded on the mould, and vacuum impregnated with resin
and afterwards cured at a temperature of 140¢XC. From the moulds standard specimens, according to ASTM
D 3039, and non-standard specimens have been obtained. The values of the mechanical strength have been as
high as 2200 MPa, but, most relevant result, the linearity of the behaviour was kept practically until the failure.
This last issue is very relevant, at it assures a behaviour independent from the previous load history of the
component. In order to evaluate the time-dependent characteristics of the material, a set of specimens will be
tested at room temperature with the 70-80% of the collapse load, in order to verify the creep behaviour of the
material. A facility is in course of manufacturing, in order to test, in operating conditions and up to stresses
leading to the rupture, rings in scale 1/10 compared to the true dimensions of the ITER rings. These tests, once
the facility will be available, will give information about the behaviour of massive structures in unidirectional
fibreglass composite in order to supply information needed for the design of the ITER rings.
Corresponding Author:
NARDI CLAUDIO
nardi@frascati.enea.it
ENEA - Via E. Fermi 45 - 00044 Frascati (Roma)
192
- E - Magnets and Power Supplies.
P2T-E-34
MEASUREMENT OF CONTACT RESISTANCE DISTRIBUTION IN
TYPICAL ITER SIZE CONDUCTOR TERMINATION
ANGHEL ALEXANDER, BRUZZONE PIERLUIGI
EPFL/CRPP Fusion Technology, CH-5232 Villigen, Switzerland
A new test facility, JORDI, dedicated to the measurement of contact resistance distribution in full-size
conductor termination at liquid helium temperature, has been designed and manufactured at CRPP Fusion
Technology in Villigen, Switzerland. The new facility is composed of a medium size liquid helium cryostat
which host a full-size termination sample, a cryogenic, parallel resistor array for providing balanced current
injection in up to 100 channels from a 10kA DC power supply (100 channels of each 100A), a pair of
conventional current leads rated at 10kA and a data acquisition system with 200 analogue input channels (100
voltage and 100 current readouts). Recently, the facility was used to characterize and qualify a first prototype
of a typical ITER joint. As opposed to the earlier attempts where the contact resistance was measured under an
unbalanced current distribution i.e. only one strand or a small group of strands were connected to the current
source while all other strands were not charged by a current, in the new facility the contact resistance is
measured under imposed balanced current distribution among all current carrying elements of the termination.
Also in the new facility all voltage drops are sensed simultaneously as opposed to the earlier measurements
where the voltage drops were measured sequentially. The results of the measurements and interpretation based
on a simple statistical, electrical network termination model are presented.
Corresponding Author:
ANGHEL ALEXANDER
anghel@psi.ch
EPFL/CRPP Fusion Technology, CH-5232 Villigen, Switzerland
193
- E - Magnets and Power Supplies.
P2T-E-35
UPDATING THE DESIGN OF THE FEEDER COMPONENTS FOR THE
ITER MAGNET SYSTEM
YOSHIDA KIYOSHI, TAKAHASHI YOSHIKAZU (1), ISONO TAKAAKI (2), MITCHELL NEIL(1)
(1) ITER, Naka Joint Work Site (2) Dept. of Fusion Engineering Research, Japan Atomic Energy Research Institute, Japan
The ITER superconducting magnet system stores energy from 40 to 50 GJ during plasma operation, and
generates an average heat load of 23.4 kW at 4.3 K to cryoplant. This heat load is removed by a primary
supercritical helium circuit with a flow of 6.0 kg/s of helium. The helium is distributed to the coil through a
complex system of 30 separate feeder lines. The feeders also contain the electrical supplies to the coil and are
integrated into the current lead transition to room temperature. The interface components between the coils and
the service facilities (power supply and cryogenic plant) consist of the in-cryostat feeders, the cryostat
feedthroughs, and the coil terminal boxes (CTBs). The layout of the in-cryostat feeders takes into
consideration routing restrictions in the cryostat and initial assembly with other Tokamak components. The
electrical break boxes on the coil surface are designed for accessibility and manual repair handling in case of
failure. The cryostat feedthroughs with S-bend boxes allow thermal contraction of the magnet system. The
forced-flow-cooled current leads in the CTBs are adapted to fit in the limited space in the building. A
conventional copper heat exchanger is used for the current lead base design although high temperature
superconducting leads could also be used if the performance would be improved relative to the base design.
Ground faults and short circuits are the main potential accidents in the superconducting feeders. A double
insulation and monitoring system together with robust mechanical containment will help prevent and /or
reduce the impact of such accidents. This paper presents the latest design concept and parameters of the feeder
components.
Corresponding Author:
YOSHIDA KIYOSHI
yoshida@naka.jaeri.go.jp
ITER, Naka Joint Work Site, 801 Mukaiyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193, Japan
194
- E - Magnets and Power Supplies.
P2T-E-36
MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR
ELECTRICAL INSTALLATIONS: COMPUTATION OF THE
BACKGROUND FIELD AND CONSEQUENCES ON THE DESIGN OF
THE ELECTRICAL DISTRIBUTION BOARDS AND CONTROL
BOARDS FOR THE ITER TOKAMAK BUILDING
BENFATTO IVONE, P.BETTINI (1) M.CAVINATO (2) A. DE LORENZI (2) D. DESIDERI (3) P. FEJOZ (4) L.
GRANDO (2) P. HERTOUT (4) J. HOURTOULE (4) D. VAN HOUTTE (4)
(1) University of Udine, Itay (2) Consorzio RFX - Association EURATOM ENEA, Padova, Italy. (3) University of Padova,
Italy (4) Association EURATOM-CEA, St Paul Lez Durance, France.
The electrical distribution boards and control boards located inside the ITER Tokamak building, are subjected
to constant, or slowly variable, magnetic field up to 70 mT, 10 mT/s. This is a very unusual environmental
condition for the components of the electrical installations, therefore very limited data are available on the
magnetic field compatibility of standard electromechanical and electronic components for low voltage
distribution boards and control boards. Being this information a necessary input for the design of the electrical
installation inside the ITER Tokamak building, EFDA placed two specific contracts dedicated to the execution
of experimental campaigns addressed to collect data on the magnetic field compatibility of standard industrial
components for electrical distribution boards and control boards. Several components of different
manufacturers have been tested and a large amount of data have been collected. The test procedures and the
results are reported in dedicated parallel papers presented at this conference by CEA and Consorzio RFX
(ENEA): the two Euratom Associations in charge of this experimental investigation. In parallel to the test
campaigns, EFDA promoted other activities, which are the subject of this paper and have been dedicated to the
following theoretical investigations: 1. to update the magnetic field map taking into account the latest ITER
reference layout of the Tokamak building; 2. to assess whether the steel embedded in building structure
produces significant effects on the magnetic field map in the areas dedicated to the installation of the electrical
distribution boards and control boards; 3. to assess the feasibility of magnetic shields, to mitigate strong
constraints in the design, manufacturing and installation of the boards. The paper reports on the computation
methods and the results obtained from the above theoretical activities. The consequences on the layout of the
electrical installations inside the ITER Tokamak building are also discussed in the paper together of the
recommendations for the design of the electrical distribution boards and control boards of the ITER Tokamak
building.
Corresponding Author:
BENFATTO IVONE
ivone.benfatto@tech.efda.org
EFDA-CSU Garching, Boltzmannstr. 2, D-85748 Garching, Germany
195
- E - Magnets and Power Supplies.
P2T-E-37
COMMISSIONING OF THE 10 POWER SUPPLIES OF THE CONTROL
COILS OF WENDELSTEIN 7-X EXPERIMENT
JAUREGI EDUARDO, T. RUMMEL, F. FÜLLENBACH
MAX-PLANCK-INSTITUT FÜR PLASMAPHYSIK, EURATOM ASSOCIATION D-17491 Greifswald, Wendelsteinstr. 1.
Germany
COMMISSIONING OF THE 10 POWER SUPPLIES OF THE CONTROL COILS OF WENDELSTEIN 7-X
EXPERIMENT E. Jauregi, D. Ganuza, I. García, J.M. Del Río, J. Lucas JEMA GJ 20160 Lasarte-Oria, Spain
T. Rummel, F. Füllenbach MAX-PLANCK-INSTITUT FÜR PLASMAPHYSIK, EURATOM
ASSOCIATION D-17491 Greifswald, Wendelsteinstr. 1. Germany In the region of Greifswald, north-east of
Germany, the so-called Wendelstein 7-X Experiment is progressing forward at the Max-Planck Institute for
Plasma Physics, IPP, to start operation into next years and become the Europe´s biggest project on “Advanced
Stellarators”. For the confinement of plasma some superconducting main field coils are placed, while for the
relative positioning 10 smaller coils, control coils, will be used. Each coil must be independently supplied, by a
high and precise direct current power supply, superposed to ac current frequency modulated. The contract for
the turn-key power supply for the control coils was awarded to the Spanish company, JEMA, which designed,
manufactured and tested the ten power supplies as well as the de-mineralised water cooling plant, main overall
control station and distribution centre. Actually the ten power supplies are already tested and commissioned
onsite, using some dummy loads as final load, and waiting for the last combine test joined to the IPP general
control system. Present abstract and future paper will refer to the results got at the partial acceptance tests as
well as mains problems occurred during commissioning, and proposed solutions. Each of the ten power
supplies should provide a controlled current compounded dc and 0-20 Hz bandwidth ac current in a range of
almost 3 kA at low voltage, 30 V, in four quadrants. Stability, precision and very low output ripple are
required. One of the most critical test was to check the output ripple measured at load for the output voltage,
specified at 1 Vpp, and for current, 1 App specified. Such a long cable distance from power converter up to
dummy load, maximum 30 meters, and the extremely low current ripple requirement, lower than 0,05%, made
very hard the test. Several testing devices and solutions were discussed to finally get a promising result of
0,02% peak to peak current ripple at full load. E
Corresponding Author:
JAUREGI EDUARDO
e.jauregi@grupojema.com
Pº DEL CIRCUITO, 10 20160 LASARTE-ORIA (SPAIN)
196
- E - Magnets and Power Supplies.
P2T-E-40
DESIGN AND COMMISSIONING OF THE NEW TOROIDAL FIELD
COIL FOR THE NATIONAL SPHERICAL TORUS EXPERIMENT
(NSTX)
NEUMEYER, CHARLES, E. BAKER, A. BROOKS, J. CHRZANOWSKI, L. DUDEK, P. HEITZENROEDER, C. JUN,
M. KALISH, T. KOZUB, R. MARSALA, R. PARSELLS, B. PAUL, H. SCHNEIDER, M. WILLIAMS, I. ZATZ
One of the key features, but also one of the most challenging aspects of the spherical torus (ST), is the
demountable toroidal field (TF) assembly which permits removal of the entire inner leg and center stack
assembly for maintenance. On February 14, 2003, following the morning test shots, the NSTX TF Inner Leg
Assembly failed at the lower Inner Leg-to-Flag joint. Analysis of the event identified shortcomings in the
structural design of the joint which led to failure after some 7200 machine pulses, with a limited number at the
full rating of 6kG. The stiffness of the structural assembly was not adequate, and repeated application of the
electromagnetic loads led to unanticipated loads in the bolts, and high local current density which eventually
led to failure. Due to the extensive damage it was not possible to recover the original TF inner leg assembly.
Furthermore, it was clear that an improved design was needed. Therefore a recovery effort was initiated,
beginning with the development of a new design. Extensive engineering resources were applied to the redesign effort to ensure a successful outcome while minimizing the time duration of the recovery period.
Extensive finite element analysis was performed to develop an understanding of the structural and thermal
behavior of the joint and to guide the development of the new design. Tests were performed to characterize the
electrical resistivity of the joint vs. pressure, the friction coefficient of the joint, the pull-out strength of the
fasteners, and other features. In addition, a mechanical prototype was exercised at the rated number of cycles
of full mechanical loads at elevated temperature, and an electrical prototype was tested at full current for the
full time duration at full mechanical loads. Based on the new design, the successful prototype testing, and the
improved instrumentation which includes a new fiber optic strain, temperature, and displacement monitoring
system, reliable operation at full rated parameters is fully anticipated. Operations at 4.5kG has been reestablished, and extensive measurements have been taken, which are presently under review. Although the
failure was unfortunate, it has led to an improved understanding of the TF joint behavior which is directly
applicable to the design of next step ST devices. This paper describes the new design, and the commissioning
of the new coil.
Corresponding Author:
NEUMEYER, CHARLES
neumeyer@pppl.gov
Princeton University Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey, 08543, USA
197
- E - Magnets and Power Supplies.
P2T-E-48
ANALYSES AND IMPLICATIONSOF V-I CHARACTERISTIC
RAINER WESCHE (1), ALEXANDER ANGHEL (1) PIERLUIGI BRUZZONE (1) PAOLA GISLON (2) LUIGI MUZZI
(2)
(1)CRPP-FT, CH-5232 Villigen-PSI, Switzerland (2)ENEA, Centro Ricerche Frascati, Via E. Fermi 45, 00044 Frascati,
Rome, Italy
Two short lengths of the NbTi cable-in-conduit conductor used to fabricate the poloidal field coil insert will be
tested in the SULTAN facility. The two conductor lengths to be investigated are distinguished by the presence
or absence of the subcable wraps. An aspect to be discussed in more detail is the voltage-current characteristic
of the two NbTi cable-in-conduit conductors. Previous results obtained for three NbTi subsize cable-in-conduit
conductors with one or two out of four petals disconnected indicate that the n factor, defined by the power law
for the electric field, is closely related to the current distribution and current transfer effects in the cable. The
presence of the subcable wraps can considerably hinder the current transfer between the individual strands in
the cable. This effect may lead to changes in the voltage-current characteristic. The implications of the results
on the poloidal field coil conductor design will be considered. Finally, the effect of the self-field on the dc
performance and the quench behaviour will be addressed. Due to the variation of the magnetic field within the
conductor cross-section the peak electric field occurring on the high field side of the conductor is much larger
than the average electric field. As a consequence the conductor quenches in the peak field region. The
measured quench currents and take-off electric fields will be simulated assuming peak-field-induced quenches.
The dc performance of the poloidal field coil NbTi conductors will be compared to previous results obtained
for five NbTi subsize cable-in-conduit conductors with parametric variations in the conductor layout.
Corresponding Author:
RAINER WESCHE (1)
rainer.wesche@psi.ch
CRPP-FT, CH-5232 Villigen-PSI, Switzerland
198
- E - Magnets and Power Supplies.
P2T-E-50
PIONEERING SUPERCONDUCTING MAGNETS IN LARGE
TOKAMAKS: EVALUATION AFTER 17 YEARS OF OPERATING
EXPERIENCE
DUCHATEAU JEAN-LUC, B. GRAVIL, M. TENA, D. HENRY, D. VAN HOUTTE
CEA Cadarache F-13108 Saint Paul Lez Durance Cedex FRANCE
Selected also for oral presentation
O2A-E-50
As a part of the Euratom program, it was decided at the beginning of the eighties that Europe will build a large
size tokamak Tore Supra, using a superconducting toroidal field magnet to demonstrate the applicability of
superconducting magnets to future fusion reactors. Part of the Physicist community was at that time reluctant
to consider this possibility, being afraid of possible difficulties in exploitation of the machine. Especially the
operation of a large refrigerator with thousands litres of 1.8 K helium was considered as completely unrealistic
and industrially impracticable. As a matter of fact, 17 years after the first plasmas, the fusion Community is
happy to have accumulated this first experience with superconducting magnets, especially at the time where
ITER is on the verge of being launched. Moreover no large fusion magnet is now considered in the world
without superconducting magnets. Far from being a burden in the exploitation, the availability of the TF
system all the day long for Plasma Physics is on the contrary of great help for the implementation of long
shots. The absence of large mechanical cycling, is also a guarantee for the good operation of the coils on the
long run. After these 17 years, the specific impact on a Tokamak operation of such a large system at low
temperature will be analysed and detailed: - Influence of daily current increase on coils temperature - Influence
of plasma shot on coils temperature - Influence of plasma disruption on coils temperature - Influence of plasma
disruption on voltage quench detection. - Specific impact of long runs (500 s) on superconducting system Behaviour of the coils during fast safety discharges (FSD) - Duration of coil recooling after night for a typical
operation day - Warming up and cooling down the coil for cryogenic system maintenance Quantitative data
will be given, of the TF for the cryogenic system and for the magnet system as well, concerning the number of
plasmas shots and the availability of the machine. The origin and the number of breakdowns or incidents will
be described, with emphasis on cyogenics to document repairs and changes on the system components.
Overall, despite the differences in design and size, the accumulated experience of 17 years of operation is a
useful tool to prepare the manufacture and the operation of the ITER magnets.
Corresponding Author:
DUCHATEAU JEAN-LUC
duchat@drfc.cad.cea.fr
CEA Cadarache F-13108 Saint Paul Lez Durance Cedex FRANCE
199
- E - Magnets and Power Supplies.
P2T-E-55
STABILITY, THERMAL EQUILIBRIUM AND DESIGN CRITERIA FOR
CABLE-IN-CONDUIT-CONDUCTORS WITH A BROAD TRANSITION
TO NORMAL STATE
NICOLAI MARTOVETSKY,
Stability in CICC (cable-in-conduit conductors) against perturbations is often associated with transient heat
removal of heat generated in the normal zone, which appeared in CICC as a result of a strong perturbation. In
such a transient condition, a simplified approach to stability calls for a sufficient amount of copper in the
strands, with sufficiently small diameter, such that the heat removal is higher than the heat generation. This
criterion is often used for design of the fusion magnets, like ITER, KSTAR and others. We show that this
criterion is not a mandatory requirement for serviceability of CICC and that CICC may work reliably at higher
current densities. In conditions of limited perturbations, a sufficient stability is provided not by a large amount
of copper and high transient heat transfer, but by a smooth transition to the normal state and ease of current
redistribution. A strand parameter space for CICC stability is proposed and discussed. The theory predictions
are compared with known experimental data on CICC that meet and do not meet this design criterion.
Corresponding Author:
NICOLAI MARTOVETSKY
martovetsky1@llnl.gov
LLNL, 7000 East Ave, L-641, Livermore, CA, 94550, USA
200
- E - Magnets and Power Supplies.
P2T-E-68
DESIGN OPTIMISATION OF THE ITER TF COIL CASE AND
STRUCTURES
MARCO FERRARI (1), PIETRO BARABASCHI (1), CORNELIS T.J. JONG (1), REINHARD K. MAIX (2),
NEIL MITCHELL (3)
(1) ITER International Team, Boltzmannstr. 2, D-85748 Garching, Germany (2) ITER VHTP, ATI Atominstitut Wien,
Stadionallee 2, A-1020 Vienna, Austria (3) ITER International Team, 801-1 Mukouyama, Naka-machi, Naka-gun, Ibarakiken, 311-0193 Japan
The basic ITER toroidal field (TF) coil case design was defined in 2001 and since then has been undergoing a
process of refinement and optimisation. The performance and major geometry of the TF coil case, which
encloses the winding pack, has remained unchanged, but significant improvements have been made, first to the
structural support, relaxing the non-destructive testing inspection levels to achieve the fatigue life, secondly to
the material options and main fabrication steps, reducing the fabrication time and cost, and thirdly to design
definition of auxiliary systems. On the structural design, the system of poloidal keys, linking the coils at top
and bottom of the central vault, has been optimised, virtually eliminating stress concentrations in the keyways.
A choice has been made between the two options for the intermediate outer intercoil structures (friction-joint
design is preferred over box), eliminating also the consideration of castings for some coil sections. The design
of the friction-joint panels and the mechanical connection to the adjacent TF coils have been optimised,
providing access to perform two-sided welding of the panels during machine assembly. The reference system
of rings for pre-compressing the inner legs of the TF coils has been confirmed as uniaxial glass fibre based on
recent R&D results and the design has been modified to allow re-tightening of the rings without the need to
remove the central solenoid. For the fabrication, the main components (the TF coil case sub-assemblies before
winding pack insertion and subsequent closure welding) have been zoned into three material grades of
austenitic stainless steel, minimising the quantity of the top grade that requires electro-slag refining to achieve
the specified performance. The final assembly of the winding pack into the case has been redefined to reduce
distortion and the risk of insulation damage due to the closure welds while minimising case-winding pack gaps
in the peak stress regions. Design of auxiliary systems has included the case cooling inside and outside,
including a thermal screen to protect the peak field region of the conductor from the maximum nuclear
heating. The paper reports the final design, including the reasons for the design choices that have been made.
Structural assessments of the critical components are also shown.
Corresponding Author:
MARCO FERRARI (1)
ferrarm@itereu.de
ITER International Team, Boltzmannstr. 2, D-85748 Garching, Germany
201
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P2T-E-73
FABRICATION OF THE PLANAR COILS FOR WENDELSTEIN 7-X
VIEBKE, HOLGER, TH. RUMMEL (1) K. RIßE (1) R. SCHROEDER (1) R. WINTER (2)
(1) Max-Planck-Institut für Plasmaphysik, Greifswald Branch, Euratom Association, Wendelsteinstraße 1, D-17491
Greifswald (2) Tesla Engineering Ltd., Water Lane, Storrington, Sussex RH20 3EA, England
WENDELSTEIN 7-X (W7-X) is a superconducting stellarator, which uses 50 non-planar coils for the main
field and 20 planar coils to modify the magnetic configuration. The coils are arranged in five modules
requiring five differently shaped non-planar and two differently shaped planar coils. The magnet system is
designed for 3 T on the plasma axis. Nominal currents of the non-planar coils are 17.6 kA against 16 kA for
the planar coils. One planar coil has an outer diameter of around 4 metres. The main elements of planar coils
are the winding packages made of a cable-in-conduit superconductor, a coil case made of stainless steel plates,
the embedding filler material, two interlayer joints to connect the double layers and a case cooling using
copper plates and stainless steel pipes. Connection of the coil to the coil support structure is performed through
forged blocks welded to the casing and bolts. Quench detection, temperature sensors and strain gauges are
installed to control operation. Manufacturing of the planar coils is contracted to the company Tesla and has to
be performed with a high accuracy to maintain the required symmetry of the magnetic configuration of W7-X.
A tolerance of 0.2 mm is allowed for the machined surfaces as compared to the CAD-model. The accuracy of
the coils is surveyed by photogrammetry. All steps of production are rigorously controlled by quality
assurance. Prior to dispatch the coils will pass a works acceptance test at Tesla thereby demonstrating helium
leak tightness, resistance against high voltage and the specified flow-resistance of the cooling channels. Prior
to delivery to Greifswald, all coils will be subject to a functional test at cryogenic temperatures at the Low
Temperature Lab of CEA. By March three coils have been delivered and one has passed successful the test at
nominal current and have shown sufficient margin. The presentation will give an overview about the status of
production and address major technical problems, which had to be solved.
Corresponding Author:
VIEBKE, HOLGER
holger.viebke@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Greifswald Branch, Euratom Association, Wendelsteinstraße 1, D-17491
Greifswald
202
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P2T-E-81
OVERVIEW OF THE DIII–D INTERNAL RESISTIVE WALL MODE
STABILIZATION POWER SUPPLY SYSTEM*
SZYMANSKI, D.D., G.L. CAMPBELL (1), W.P. CARY (1), R. HATCHER (2), G.L. JACKSON (1), A.G.
KELLMAN (1), A. NAGY (2), AND C.J. PAWLEY (1)
(1) General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (2) Princeton Plasma Physics Laboratory,
Princeton, New Jersey
With the recent installation in the DIII-D Tokamak of internal resistive wall mode (RWM) stabilization coils
(I-Coils), upgrades to the existing RWM and error field correcting power supply systems were necessary. The
new I-Coil system is comprised of 12 individual low inductance magnetic field coils that can be rearranged in
multiple configurations with the main purpose of providing feedback stabilization up to the ideal wall beta
limit without the need for strong plasma rotation. This paper will discuss the power supply system upgrades
needed to drive up to 5 kA in these low inductance coils. The power supply system is now comprised of (5)
300 Vdc, 5-7 kA pulse rated power supplies which can either be connected directly to magnetic coils or else
provide input power to (4) 300 Vdc, 5 kA pulse rated switching power amplifiers (SPAs). The SPA actuators,
when connected to the I-Coils provide maximum current from dc to 300 Hz and can operate up to 2 kHz at
reduced current, limited by the inductance of the I-Coils and their cable feeds. In some experimental scenarios
faster response with lower phase shift is required than can be provided by the SPAs. In this case, high power
audio amplifiers will be installed. We will present the details of the upgraded power system including
instrumentation, data acquisition, multiple SPAs powered by a single dc supply, a versatile patch panel, and
low inductance cabling. In addition, the design of audio amplifiers will also be discussed. *Work was
supported by the U.S. Department of Energy under DE-FC02-04ER54698 and DE-AC02-76CH03073.
Corresponding Author:
SZYMANSKI, D.D.
szymanski@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
203
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P2T-E-95
THE EUROPEAN DEVELOPMENT OF A FULL SCALE SWITCHING
UNIT FOR THE ITER SWITCHING AND DISCHARGING NETWORKS
BONICELLI TULLIO, A. DE LORENZI (1) E. GAIO (1) D. HRABAL (2) F. MILANI(1) R. PIOVAN (1) E. SACHS
(2) E. SALPIETRO (4) S. SHAW (3) V. TOIGO(1) L. ZANOTTO (1)
(1) Consorzio ENEA-RFX, Italy (2) FEAG, Germany (3) UKAEA, UK (4) EFDA-CSU Garching
Selected also for oral presentation
O2A-E-95
The ITER coil power supply systems are provided with Discharging Networks whose main purpose is to
dissipate in resistors the magnetic energy stored in the super conductive coils when a quench is detected.
Similarly, Switching Networks are series connected to some of the poloidal coils and to the central solenoid to
produce the required loop voltage during plasma start-up. These actions are performed diverting the current
flowing in a switch into a resistor connected in parallel. The European Fusion Programme included since the
mid-90’s the development of a full scale, full rating switching unit to be used as centre-piece of the ITER
Commutating Units. The main rating of such units are: steady state and breaking current of 70 kA, peak
withstand current 250 kA, rated voltage 17.5 kV rms, recovery voltage 24 kV. Since the current carrying
capability in steady state of a vacuum circuit breaker (VCB) is far from been sufficient for the steady state
operation, the combined operation of a mechanical by-pass switch (BPS), rated for the continuous current, and
a vacuum circuit breaker has been proposed and developed. During the first phase of the activity (years 19951999), the VCB and the BPS were individually characterised and tested at their rated performances, including
the execution of full scale life testing. An important limitation on the maximum I2t in the VCB before opening
was identified. The second phase, concluded at the beginning of 2004, was devoted to check for the first time
the combined operation of the by-pass switch and the vacuum circuit breaker up to the full performances.
Some minor improvements of the switches were also tested, coming from the results of the first phase. In the
paper the successful results of the type and life tests will be presented, including the operation in absence of
the saturable reactor series connected in the ITER reference design, which could yield some cost and space
savings. Interruption tests at low current and full counterpulse capacitor voltage were also successfully
performed. The testing included an extensive and novel characterisation of the interrupting capability of the
vacuum circuit breaker in the presence of an external magnetic field as in the actual location of installation in
ITER, where a stray magnetic field of up to 25 mT will be present.
Corresponding Author:
BONICELLI TULLIO
tullio.bonicelli@tech.efda.org
EFDA-CSU Garching, Boltzmannstr. 2, D-85748 Garching, Germany
204
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P2T-E-97
MECHANICAL PERFORMANCE OF MAGNET INSULATION
MATERIALS FABRICATED BY THE “INSULATE-WIND-AND-REACT “
TECHNIQUE*
DR. HUMER KARL, KARIN BITTNER-ROHRHOFER (1) KARL HUMER (1) HARALD FILLUNGER (1) REINHARD
K. MAIX (1) HARALD W. WEBER (1)
(1) Atomic Institute of the Austrian Univ. Stadionallee 2 1020 Vienna Austria
Usually, superconducting magnet coils are fabricated according to the standard “Wind-React-Insulate-andTransfer” technique, where the superconductor is wound and heat treated first, before the insulation is applied
and the coil vacuum impregnated with epoxy. In order to simplify the manufacturing process of such coils and
to lower the costs, an alternative procedure, the “Insulate-Wind-and-React” technique can be used. In this case,
the superconductor is insulated first, followed by winding, heat treatment and impregnation of the coil, i.e. the
“transfer”-step can be avoided. Such an insulation system fabricated by European industry (Ansaldo, Italy) has
been investigated. It consists of a two-dimensional R-glass-fiber reinforcement heat treated at 650 C and
impregnated afterwards with epoxy. In order to characterize the mechanical material performance, both tensile
and short-beam shear (SBS) tests were carried out at 77 K. The ultimate tensile strength is about 500 and 250
MPa parallel and perpendicular to the glass fibers, respectively. Furthermore, tension-tension fatigue tests
were done to simulate dynamic load conditions caused by the Lorentz forces. In addition, a set of SBS
samples, irradiated to a fast neutron fluence of 1x1022 m-2 (E>0.1 MeV), is in order to check for material
degradation induced by radiation. *This work has been carried out within the association EURATOM-OEAW.
Corresponding Author:
DR. HUMER KARL
khumer@ati.ac.at
Atomic Institute of the Austrian Universities, Stadionallee 2, 1020 Vienna, Austria
205
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P2T-E-99
INFLUENCE OF PARAMETER VARIATIONS ON THE FATIGUE
BEHAVIOR OF MAGNET INSULATION SYSTEMS
PROF. WEBER HARALD. W., HUMER KARL (1) WEBER HARALD W.(1)
(1)= TU-Wien/Atominstitut der Österreichischen Universitäten Stadionallee 2 A-1020 Wien AUSTRIA
The reliable application of glass-fiber reinforced plastics as insulation materials for fusion magnet coils (e.g.
the Toroidal Field Coils of ITER) requires the full characterization of their mechanical performance under
ITER-relevant conditions. One of the common methods to test the material’s response under dynamic load is
the tension-tension fatigue procedure. This test can be used to simulate the pulsed tokamak operation of the
ITER coils over a lifetime of more than 20 years. Furthermore, it provids information on the maximum tensile
or shear stress in the ITER-relevant region of 104-105 cycles. In order to simulate the operation conditions of
ITER as closely as possible, several fatigue parameters can be set in the test programme, e.g., the minimum-topeak stress ratio R and the frequency n of the sinusoidal load function. Further, the fatigue process can run
under load or displacement control. All of these parameters may influence the mechanical response of an
insulation system under cyclic load. Therefore, it is highly desirable to investigate the influence of test
parameter variations on the measured stress-lifetime diagrams. The investigations were performed at 77 K
using an industrial glass-fiber reinforced composite impregnated with epoxy resin. For both the load and the
displacement controlled mode, R-values of 0.1-0.5 and frequencies of 5-20 Hz were chosen. The results will
be discussed and compared with respect to ITER-relevant operation conditions. *This work has been carried
out within the association EURATOM-OEAW.
Corresponding Author:
PROF. WEBER HARALD. W.
weber@ati.ac.at
TU-Wien/Atominstitut der Österreichischen Universitäten, Stadionallee 2, A-1020 Wien, AUSTRIA
206
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P2T-E-101
MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR
ELECTRICAL INSTALLATIONS: TESTS ON PROGRAMMABLE
LOGICAL CONTROLLERS AND OTHER ELECTRONIC DEVICES
HOURTOULE JOEL, D. VAN HOUTTE P. FEJOZ P. HERTOUT
The electric switchboards installed in the ITER Tokamak building can be subjected to a static or slightly
variable magnetic field induced by the ITER coils, of a value that can reach 70 mT. This environment is really
particular and doesn’t find itself in any industrial facility. There are no experiments on this subject and the
components manufacturing standards don’t take into account this aspect of magnetic field compatibility. An
experimental test campaign has been launched by EFDA, in collaboration with CEA and Consorzio RFX, in
order to find the operational limits of the components employed. In this collaboration, CEA took the
responsibility for the tests of the electronic components, of the signals conditioning and command control
units. For these tests, a test bench has been developed in CADARACHE, composed by a solenoid and a remote
control power supply. The choice of the components has been carried out in collaboration with manufacturers,
by choosing in middle range material and having a lifespan of at least five years. Particular tests procedures
were applied, strongly inspired by the standards in force for the tests in the presence of alternate magnetic
field. The tests showed that all the components are more or less sensitive to this type of environment. The
observed effects vary from the simple temporary dysfunction until the total destruction of internal electronic
component. As expected, the most sensitive components were those presenting a ferromagnetic part, such as
the relays or galvanic transformers. Moreover, it was shown the importance of the direction of field. The
results record the limits in each position, but retain only the most unfavourable position limit. For the signal
conditioning units, a significant increase in consumption was observed. The limits of such components are at
about 30 mT. For the command control systems (PLC and peripheral) the limits were found around 40 mT.
The most sensitive components are relays, which show operational limits according to their position in the
field, below of 20 mT. These first results, that need to be refined, shall to be taken into account, not only in the
design of the electric distribution boards, but also for all the sets of measurements that will be installed in the
TOKAMAK building and will be subjected to a significant magnetic field.
Corresponding Author:
HOURTOULE JOEL
joel.hourtoule@cea.fr
DRFC/ STEP BT502 13115 SAINT PAUL LEZ DURANCE
207
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P2T-E-105
DESIGN, FABRICATION AND INSTALLATION OF CRYOGENIC
TARGET SYSTEM FOR 14 MEV NEUTRON IRRADIATION
NISHIMURA ARATA, HISHINUMA YOSHIMITSU (1) TANAKA TERUYA (1) MUROGA TAKEO (1) NISHIJIMA
SHIGEHIRO (2) SHINDO YASUHIDE (3) TAKEUCHI TAKAO (4) OCHIAI KENTAROU (5) NISHITANI TAKEO (5)
OKUNO KIYOSHI (6)
(1) NIFS, Gifu 509-5292 Japan (2) Osaka Univ., Osaka 565-0871 Japan (3) Tohoku Univ., Miyagi 980-8579 Japan (4)
NIMS, Tsukuba, Ibaraki 305-0047 Japan (5) JAERI, Tokaimura, Ibaraki 319-1195 Japan (6) JAERI, Nakamachi, Ibaraki
311-0193 Japan
The design of the International Thermonuclear Experimental Reactor (ITER) has been progressed and the
neutron streaming is clarified analytically. The hard streaming is expected around NBI ports and it will cause
irradiation on superconducting magnets. Since the irradiation spectrum is different depending on the location,
the effect of pure 14 MeV neutrons on materials is planed to be investigated to clarify the change of
mechanical and electrical properties of the magnet materials. A cryogenic target system has been installed in
Fusion Neutronics Source (FNS) at Japan Atomic Energy Research Institute (JAERI) under collaboration
between Universities, National Institutes and JAERI and makes it possible to perform the electrical
measurement at cryogenic temperature without warming up the samples. Deuterium is accelerated to around
350 keV and collides with tritium absorbed in rotating target plate, resulting in D-T reaction which generates
14 MeV neutrons. The neutron flux depends on the distance (r) from the collision point and decreases as a
function of 1/r2. The cryogenic target will be located at about 10 mm far from the D-T reaction point and be
kept at 4.5 K by a small refrigeration system whose capacity is 0.5 W at 4.2 K. The compressor and data
acquisition system are installed in the other room to reduce the neutron irradiation and operated automatically.
Samples of Nb3Sn, NbTi, Nb3Al and pure copper wires will be attached on the target plate and be irradiated
up to the fluence of 1016 n/cm2 at cryogenic temperature. According to the reference data, it is expected that
Nb3Sn and NbTi will show no change in such neutron fluence. However, the electric resistance of pure copper
will be increased and it is important to clarify the change in resistance at cryogenic temperature, for the pure
copper is commonly used as a stabilizer for the superconducting strand. At the same time, organic materials
will be irradiated in a room temperature space together with glass fiber reinforced plastics (GFRP) and the
mechanism of the decomposition and the interlaminar shear strength will be discussed.
Corresponding Author:
NISHIMURA ARATA
nishi-a@nifs.ac.jp
Fusion Engineering Research Center, National Institute for Fusion Science, Oroshi 322-6, Toki, Gifu 509-5292
Japan
208
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P2T-E-106
THE EUROPEAN NB3SN ADVANCED STRAND DEVELOPMENT
PROGRAMME
VOSTNER ALEXANDER, E. SALPIETRO (1)
(1) EFDA Close Support Unit - Garching, Boltzmannstr. 2, 85748 Garching, Germany
Significant progress in the field of Nb3Sn strand manufacture has been made over the last few years. Strands
relevant for fusion with high critical current densities and moderate hysteresis losses have been developed and
already produced on industrial scale for the KSTAR project. Based on these achievements EFDA CSU –
Garching has launched a Nb3Sn strand development and procurement action inside Europe in order to assess
the current status of the Nb3Sn strand production capability. All six addressed companies replied positively to
our strand R&D programme which includes the three major Nb3Sn production techniques namely the bronze,
internal-tin and powder-in-tube (PIT) route. According to the strand requirements for the ITER TF conductor a
critical current density of 800 A/mm2 (at 12 T, 4.2 K and 10 µV/m) and overall strand hysteresis losses below
500 kJ/m3 have been specified as the minimum guaranteed strand performance. The second major objective of
this programme is to motivate the strand manufacturers in utilising the technical advances to develop and
design advanced state-of-the-art Nb3Sn strands optimised for the ITER conductor. For this purpose, a target
critical current density of 1100 A/mm2 has been added to the specification. This paper describes the strategy
behind the strand development programme, the actual status of the strand production as well as first
preliminary results obtained from the strand suppliers.
Corresponding Author:
VOSTNER ALEXANDER
alexander.vostner@tech.efda.org
EFDA Close Support Unit - Garching, Boltzmannstr. 2, 85748 Garching, Germany
209
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P2T-E-111
DESIGN AND DEVELOPMENT OF THE POWER SUPPLY SYSTEM
FOR HL-2A TOKAMAK
YAO LIEYING, XUAN WEIMIN LI HUAJUN CHEN YUHONG BU MINGNAN SHAO KUEI HU HAOTIAN MAO
XIAOHUI WANG SHUJIN REN JUQIAN
Southwestern Institute of Physics, P.O. Box 432,Chengdu,Sichuan,610041, P.R.China
The HL-2A is the first divertor tokamak in China. Its construction is based on the main components of
ASDEX from IPP and an entirely new power supply system is required to power its magnetic field coils and
the plasma heating systems. The most important parameters of the HL-2A are toroidal field of 2.8T, plasma
current of 480 kA with a flat top of 5s. Thus, the peak power required is 300MVA and the energy content is
about 1200MJ per shot. Three flywheel motor-generators (MG) are used to transfer the power and energy from
the HV grid. To get sufficient released energy, two identical existing MG have been modified by replacing
original flywheel to a big one. Raising the maximum speed and increasing the speed drop of the total shaft are
the other ways adopted to increase the released energy. After modification, the maximum apparent power for
each generator can increase to 90MVA from 80MVA and released energy can rise to 500MJ from 100MJ.Two
modified MG are used to power the toroidal field coils via a 12 pulse diode rectifier. Another MG with output
power of 125MVA is used to power the poloidal field system with transformers and thyristor rectifiers. In
order to check the initial design and optimize the feedback control system parameters, all the important parts
of the power supply system have been simulated with EMTP code. A digital trigger circuit with the precision
of 0.04 degree and a reliable protection system are developed to ensure the performances of the power supply.
The feedback control of the plasma current and position were worked successfully both in limiter and divertor
operations in 2003. The primary tests show that the design and development of the HL-2A power supply
system basically meet the requirements of the operation of the HL-2A.
Corresponding Author:
YAO LIEYING
yaoly@swip.ac.cn
Southwestern Institute of Physics, P.O. Box 432,Chengdu,Sichuan,610041, P.R.China
210
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P2T-E-112
THE ITER THERMAL SHIELDS FOR THE MAGNET SYSTEM: DESIGN
EVOLUTION AND ANALYSIS
BYKOV VICTOR, YU.KRASIKOV(2), S.GRIGORIEV(2), V.KOMAROV(2), V.KRYLOV(2), A.LABUSOV(2),
V.PYRJAEV(2), S. CHIOCCHIO(3), V.SMIRNOV(2), V.SORIN(2), V.TANCHUK(2)
(2)D.V. Efremov Scientific Research Institute, St.Petersburg 196641, Russia (3)ITER IT, Boltzman Str 2, 85748 Garching,
Germany
The ITER thermal shield (TS) system is designed as a continuous barrier, that reduces by over two orders of
magnitude the heat loads transferred by thermal radiation and conduction from warm components to the
components and structures that operate at 4.5K. Active cooling of the TS by 80K gaseous He, and the
provision of silver coated surfaces with low emissivity facilitate the removal of the residual 6 kW heat load on
the magnet system during normal conditions by the cryoplant of reasonable capacity. By making the TS
components independent toroidally continuous structures the number of TS supports and hence conductive
heat transfer are minimised. After step-by-step modifications the ITER TS consists of three main
subcomponents: (1) the central TS, which comprises the vacuum vessel TS (VVTS) around the hot vacuum
vessel, central cryostat and transition TS; (2) the upper cryostat TS suspended from the cryostat lid; and (3) the
lower cryostat TS supported on the cryostat floor. The TS system also includes the support TS side panels, that
block heat loads to and from the magnet gravity supports (MGS) and thermal anchors in the MGS. The
efficiency of the TS system depends strongly on the interface between its components, therefore minimisation
of the number of the TS components and reduction of heat loads through interfaces is the main approach of the
design evolution. The requirements for access to magnet components for repair cannot be excluded for the
ITER machine. Locating the cryostat TS just outside the PF coils provides the required space for in-cryostat
repair activity outside delicate TS surfaces, while the incorporation of removable TS panels and modification
of the outboard Central TS support make access to the TS/Magnet interspace for assembly and disassembly
relatively easy. The modern design of the VVTS with extruded, profiled cooling pipes hidden in between
double panels improves surface smoothness and thermal efficiency of the structure. Complex tube tracing
avoids twisting of the pipes. This paper presents the rationale for the TS design evolution since 2002. Details
of the recent modifications that affect the TS cooling panels, the Central TS ports and support system, interface
labyrinths and TS structural joints as well as the modern results of thermal-hydraulic, thermal, seismic, static
and dynamic structural analyses, that involve submodeling and substructuring finite element analysis
techniques, are presented.
Corresponding Author:
BYKOV VICTOR
bykovv@itergps.naka.jaeri.go.jp
ITER IT, 801 Mukouyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan
211
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P2T-E-121
QUALITY ASSURANCE PROCEDURES IN THE EAST MAGNETS
MANUFACTURING PROCESS
CHEN SIYUE, GAO DAMING, YU JIE, WU JIEFENG, ZHANG PIN, TAO YUMING
Institute of Plasma Physics, Chines Academy of Sciences. Hefei, Anhui, China
EAST is a full super conducting Tokamak being constructed in Hefei, China. At the beginning of next year it
will have been assembled. Three toroidal field (TF) magnets have been made so far and the first one has
passed all the examination. The operating current of TF magnet system is 14.3 KA and its toroidal field is 3.5
T. All the coils are winded by Cable-in-conduit conductors (CICC). The mechanical property, position and
dimension precision, the electric, cooling and vacuum performance of the magnets are guaranyeed by quality
assurance procedures in fabricating process. ISO9001 quality assurance model is applied in the design and
fabricating process. The priority of quality control is design and manufacturing personnel training. Suitable
machining and testing tool and clean environment are necessary. The magnets are manufactured on four
production lines, namely CICC jacketing line, coil winding line, vacuum pressure impregnation line and
mechanical machining line. Every production line has its detailed quality plan. It describes the ways and the
standards of acceptance inspection and testing, the detailed techniques, parameters and the testing standard of
every working procedure, the track recording forms, the testing result forms, the identifying ways. Because
these magnets are not standard products, the technology and testing standard are based on a great deal of
experiments. To ensure high reliability, sometimes many testing ways are applied in one working procedure.
Corresponding Author:
CHEN SIYUE
siyue@ipp.ac.cn
Institute of Plasma Physics, Chinese Academy of Sciences. Hefei, Anhui, China
212
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P2T-E-126
THYRISTOR CROWBAR SYSTEM FOR THE HIGH CURRENT POWER
SUPPLIES OF ASDEX UPGRADE
CLAUS-PETER KÄSEMANN (1), LOU VAN LIESHOUT (2) MICHEL HUART (1) CHRISTOF SIHLER (1)
(1) Max-Planck-Institut für Plasmaphysik (IPP), EURATOM Association, Boltzmannstrasse 2, D-85748 Garching, Germany
(2) Imtech Vonk BV, Modem 30, NL 7741 MJ Coevorden, The Netherlands
The ohmic heating system and the poloidal field coils of ASDEX Upgrade (AUG) are supplied by 15 thyristor
converter units with an installed apparent power of 600 MVA. A Thyristor Crowbar System (TCS) consisting
of 15 units (TCU) was designed, installed and commissioned. These will be used for protecting the thyristor
converters against DC overvoltage arising from abnormal operations and resulting damages caused by the
large energy stored in the AUG magnet coils. The TCS has to fulfil three main objectives: Reliability intervention in case of overvoltage but no tripping due to false alarms; Modularity - independent operation of
all units; Flexibility - selection of triggering voltage taking account of the different DC system voltages. Each
TCU is connected to the DC output terminals of one of the thyristor converters. There are three types of TCU,
characterised by their DC rated voltage, namely 2400 V, 1500 V and 500 V. The DC rated current is 45 kA.
The TCU is triggered by a DC overvoltage of either polarity and suitable to carry DC current of both
polarities. In case of overvoltage the trigger circuit fires a thyristor that transfers the current from the converter
resp. load coil to a resistor where the energy is dumped. Each dump resistor is composed of series connected
resistor banks that are each characterised by a rated pulsed energy of 5 MJ and a nominal resistance of 25
mOhm. To increase reliability each TCU comprises two modules parallel connected that include their own
overvoltage detection, trigger circuit, dump resistor and a ´cross-firing´ between the two modules. The triggerlevel is chosen half way between the DC rated voltage and the thyristor blocking voltage. If the converter
configuration is changed the trigger voltage can easily be adapted by changing the overvoltage detection
board. Each TCU includes its own instrumentation and interlock to ensure all necessary interfaces with the
AUG control system and the thyristor converters interlock system. This paper describes the design and testing
of the Thyristor Crowbar System representing the DC converter overvoltage protection system. It will present
the layout, analyse the results of measurements obtained during commissioning, compare them to the
calculated (design) values and report on the first experience on the AUG coils improving the safety of the
equipment.
Corresponding Author:
CLAUS-PETER KÄSEMANN (1)
C.P.Kaesemann@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik (IPP), EURATOM Association, Boltzmannstrasse 2, D-85748 Garching,
Germany
213
- E - Magnets and Power Supplies.
P2T-E-186
OPTIMIZATION OF THE POWER SUPPLY FOR A HELIAS REACTOR
SUPERCONDUCTING COIL SYSTEM
HARMEYER, EWALD (1), WIECZOREK, ANDREAS (2) WOBIG, HORST (1)
(1) Max-Planck-Institut fur Plasmaphysik, EURATOM Association, D-17491 Greifswald, Germany. (2) FH-University of
Applied Science, D-93049 Regensburg, Germany.
For magnetic confinement of a hot fusion plasma Stellarator magnetic field configurations of the Helias type
have been developed. A Helias Reactor coil system with 4 field periods and a major radius of 18m is applied.
This coil system comprises 40 modular coils in total, 10 coils per field period. The electrical circuit consists of
5 coil groups, each of them with 8 equally-shaped coils connected electrically in series. These 5 coil systems
will be powered individually by 5 power supplies of the thyristor type. The power supply units must generate
currents up to 40kA to achieve a magnetic field of 5T on axis resulting in total stored magnetic energy of about
100GJ in the coil system. All systems will be powered direct from the medium voltage utility interface for
auxiliary systems. The grid is loaded by the operation of the line commutated converters with reactive power
and harmonics. Because of high power levels associated with this application, it is important to reduce the
harmonic currents generated on the ac side of the converter. This is accomplished by means of a 12-pulse
converter operation. If several converters are connected to the same supply mains, they will affect one another
through the commutation notches. Operation directly in parallel is not possible, they must therefore be
decoupled by transformer inductances. A power supply system for feeding the superconducting coils of the
Helias reactor has been investigated. This multiconverter supply system has been optimized, in view of low
losses in the components and only little impact to the power grid. The design of the optimized multiconverter
supply system was studied by means of computer simulations, using the SIMPLORER code. The influence of
the passive structures on operation of the power supply system was taken into account. The influence of
induced eddy currents in the coil structure during transient processes are transformed into electric network
analyses by means of the Finite Element Network (NET) method. This approximation allows the investigation
of the entire coil system including power supplies and passive structures.
Corresponding Author:
HARMEYER, EWALD (1)
ebh@ipp.mpg.de
Max-Planck-Institut fur Plasmaphysik, EURATOM-Association, D-17491 Greifswald, Germany
214
- E - Magnets and Power Supplies.
P2T-E-198
QUENCH CURRENT MEASUREMENT AND PERFORMANCE
EVALUATION OF THE EAST TOROIDAL FIELD COILS
WENG PEIDE, Z.M.CHEN, Y.WU, Y.N.PAN, W.G.CHEN, Z.R.OUYANG, H.Y.BAI, X.N.LIU, P.FU, L.W.XUE,
Y.F.TAN
Institute of Plasma Physics Chinese Academy of Sciences, p.o.box 1126, Hefei,Anhui 230031,China
Selected also for oral presentation
O2A-E-198
EAST device (original name is HT-7U) is a superconducting tokamak constructing in Institute of Plasma
Physics Chinese Academy of Sciences. The TF magnet system of the device is consisting of 16 D shape TF
coils made of NbTi Cable In Conduit Conductor. The nominal operating current of the coil is 14.3 kA. It is
planed to test all of the 16 TF coils this year. The test program included a number of items such as cryogenichydraulic property, electro-magnetic property and quench current measurement at different temperature. Up to
now, 11 TF coils have been tested in our test facility. Each coil was cooled down up to 4.5 K and charged to
16 kA, the magnet field on the coil, internal joint resistance of the coil, mass flow rate and pressure drop of
each cooling channel, were measured at same time. After that, the quench current of TF coil was tested, due to
limitation of power supply, we have to use higher temperature, the Helium temperature were increased to more
than 7.5 K and the coil excitation were performed again till the coil quench. The measurement results and coil
performance evaluation are presented in this paper.
Corresponding Author:
WENG PEIDE
pdweng@mail.ipp.ac.cn
Institute of Plasma Physics Chinese Academy of Sciences, p.o.box 1126, Hefei,Anhui,230031,China
215
- E - Magnets and Power Supplies.
P2T-E-209
THERMAL AND STRUCTURAL ANALYSIS OF THE W7-X MAGNET
HEAT RADIATION SHIELD
NAGEL, MICHAEL, SEONG YEUB SHIM FELIX SCHAUER
The magnet system of the fusion experiment WENDELSTEIN 7-X comprises 70 superconducting coils. In
order to reduce the heat load on the coils, in addition to high vacuum an efficient thermal insulation is required
which basically covers the outside of the plasma vessel, the inside of the outer vessel, and the outside of the
port walls. The insulation consists of multi-layer insulation (MLI) and a thermal shield which is cooled by
gaseous helium. Detail engineering of the plasma vessel insulation has been finished, and its production has
started. The paper presents the mechanical design as well as the cooling concept of the shields, and shows the
resulting temperature distributions for different design options. Calculations are based on finite element
models of the outer and plasma vessel as well as the port shields. The shields are all subdivided into panels
which are described with thermal SHELL elements. FLUID PIPE elements are used to model helium in the
cooling tubes. Heat load on the panels in normal operation is assumed to be 6 W/m2, with a uniform
distribution. The helium is warmed up from about 40 K at the inlet to around 70 K at the outlet. Structural
analysis of the thermal shield is carried out in order to define its mechanical strength as well as the required
number and positions of the supports. Main loads are electromagnetic forces resulting from a rapid shut down
of the magnet system, the weight of the shield, and treading on during cryostat assembly works. The eddy
current forces induced in the shields during a rapid shut down are calculated in a two step procedure. First the
magnetic field and the corresponding vector potential data are calculated using the well known EFFI code
based on the Biot-Savart law. In the second step, the vector potential is used as input parameter for calculating
the induced currents on the shields with the finite element code ANSYS. For the transient calculation, an
exponential decay of the magnetic field is assumed. That way only a finite element model of the shields is
required, reducing the model size and required calculation time. Results for different options are discussed. It
is shown that the chosen shield design fulfils the requirements.
Corresponding Author:
NAGEL, MICHAEL
Michael.Nagel@ipp.mpg.de
Max- Planck- Institut fuer Plasmaphysik, Wendelsteinstrasse 1, D- 17491 Greifswald, Germany
216
- E - Magnets and Power Supplies.
P2T-E-212
FILAMENT POWER SUPPLY (AC TO AC CONVERTER) FOR LONG
PULSE NEUTRAL BEAM INJECTOR OF SST-1
P.J. PATEL, O.RAJA, N.P.SINGH, V.SHARMA, U.K.BARUAH ,S.K.MATTOO AND NBI TEAM INSTITUTE FOR
PLASMA RESEARCH, GANDHINAGAR, INDIA – 382428
Filament Ion Sources used for Neutral Beam Injectors use AC heating. For long pulse operation, AC filament
heating power is advantageous. To minimize the ripple on plasma density, number of phases of the heater
supply is usually made large. This paper presents the design and performance of AC to AC converters for
filament power supply of the ion source for the long pulse Neutral Beam Injector of SST-1 (Steady-state
Superconducting Tokamak-1). The input is from the utility mains, the input stage unity power factor controller
circuit maintains the total harmonic distortion of line within 5% and power factor at unity. A PWM inverter
along with output filters generates the output. Three phase AC output is controllable from 40-400V(rms),
400Hz, sinusoidal, at 7.0 kVA (max.). The secondary from a 3-phase step down transformer (ratio 22:1) placed
at the output of the converter is connected to one filament. The stability at the output is 0.1 % with variations
in input line or fluctuations in the load. Overall efficiency is approximately 90%. Eight power supplies, all
capable of being synchronized with an external trigger pulse are used to generate a 24-phase filament heating
system for the ion source. The PWM generation carrier waveform is generated by a digital scheme, with a start
trigger for each cycle. For synchronized operation of all converters, the output can be locked to a phase
reference input TTL pulse train, whereby a 24-phase (or any user defined) system is realized within 1.0 degree
(electrical) accuracy. The control is matched to meet the filament temperature stability with compensation for
actual current in the filament. The design uses novel power topology, a combination of a high frequency
inverter and a front-end power factor controller for each phase. IGBTs are used for both the input unity power
factor controller and inverter sections.
Corresponding Author:
P.J. PATEL
paresh@ipr.res.in
Institute for Plasma Research,Bhat, Dist. Gandhinagar-382428
217
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P2T-E-219
TRANSIENT ELECTRICAL BEHAVIOUR OF THE ITER TF COILS
DURING FAST DISCHARGE AND TWO FAULT CASES
STEFAN FINK (1), TULLIO BONICELLI (3) WALTER H. FIETZ (1) AMIR M. MIRI (2) XIANGMING QUAN (2)
ALBERT ULBRICHT (1)
(1) Forschungszentrum Karlsruhe, ITP, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany (2)
Universität Karlsruhe, IEH, Kaiserstr. 12, 76128 Karlsruhe, Germany (3) EFDA-CSU Garching, Boltzmannstr. 2, 85748
Garching, Germany
Insulation faults are regarded from the ITER International Team as the most probable cause of magnet failure.
Considering the difficulties involved in the replacement of a TF coil in the ITER magnet system and the
different problems occurring during high voltage tests of the ITER model coils further improvements in
several aspects of high voltage technology for the realisation of the ITER magnets are indispensable. One of
these aspects is the consideration of the transient electrical behaviour because it is well known that fast
changes of voltages (e. g. lightning and switching impulses) may cause a non linear voltage distribution on the
coil turns and possibly excite resonances within a large coil. Such high voltage stress can cause local
overloading and irreversible destruction of the insulation system. This paper will present the calculation of the
terminal voltages within the ITER TF coil system and the voltage stress of the three insulation types (ground,
radial plate and conductor insulation) within an individual ITER TF coil for the fast discharge and two fault
cases. An electrical network model for the ITER TF coil was developed and simulated with the code PSpice.
The internal inductances and capacitances as well as the capacitances to ground for the establishment of this
network model were determined. Skin and proximity effect as well as the damping caused by eddy currents in
the stainless steel radial plates, in which the conductor is embedded, were calculated by the FEM code
Maxwell. For the complete TF circuit, composed of 18 TF coils and 9 fast discharge units, an additional
network model was set up and implemented with the code PSpice. Due to the large size of the individual ITER
TF magnets the resonance frequency is lower than for the TF model coil. It was also determined that the three
types of insulation within a single TF coil are stressed with a nonlinear voltage distribution under a fast
discharge condition. The non linear voltage distribution is enhanced in case of fast excitations applied in
consequence of ground faults. Therefore insulation coordination and test voltages have to be defined in
consideration of the stresses caused by fast discharges and applicable and realistic fault cases to ensure a
reliable operation during the foreseen ITER lifetime. Hence some proposals for the high voltage test
procedures will be discussed based on the calculated voltage stress and the experiences gained during the
ITER TF Model Coil test.
Corresponding Author:
STEFAN FINK (1)
stefan.fink@itp.fzk.de
Forschungszentrum Karlsruhe - ITP, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen
218
- E - Magnets and Power Supplies.
P2T-E-223
STUDIES ON THE BEHAVIOR OF MULTISECONDARY
TRANSFORMERS USED FOR REGULATED HV POWER SUPPLIES
N P SINGH, U K BARUAH, P J PATEL, S K MATTOO & NBI TEAM
Multisecondary winding transformers have been in use for the input stage of modular High voltage power
supplies for a long time now. These power supplies have been widely used for Neutral Beam Injectors, RF&
Microwave devices. This design is also used in pulsed electric field applications. For tokamak experiments of
steady state operation, suitability of these transformers needs to be analysed. Secondary windings of
multisecondary transformer are displaced axially on the core in the form of stacks and each stack comprises of
number of coils permitted by winding insulation. Combination of non-uniform and uniform insulation provides
required sequential voltage gradient among multisecondaries and to the ground. The secondary windings are
designed to achieve identical transformation ratio and impedances. Against each stack of the secondary, the
primary winding is designed for ampere-turns to balance short-circuit forces during a fault. In a developed and
tested design of a transformer, conventional bushing for large number of terminals is replaced by a compact
resin cast bushing plate with embedded terminations. Accommodating large numbers of secondaries (~20 or
more) in a compact transformer design leads to formations of various capacitances among secondaries (few
tens of nanofarads) and to the ground (few hundred picofarads). As the output voltage is built up by switching
of a rectifier and IGBT at the secondary, the potential of the winding is lifted to the DC potential. As switching
of different stages progresses, the winding potential fluctuates at submultiples of the switching frequency of
the IGBT. The winding stray capacitances charges and discharges, consequently, a high frequency leakage
current is induced. Effect of these stray capacitances has been observed on the output performance of the
power supply, it is likely to determine suitability of the transformers for long term operation. This paper
presents a systematic study of the interactions of the stray capacitances on the behavior of the power supply.
Results from simulation and experiments on a pair of 300kVA, 415/330V (20 secondary), 30kVDC isolation
transformers used for generating 14kVDC, 35Amp output are presented. Possible effects on the performance
of the transformer and the design considerations are discussed. Additional dielectric losses, voltage swings at
the switching frequency of the semiconductor devices and the observed effects of these factors are presented.
Corresponding Author:
N P SINGH
npsingh@ipr.res.in
NBI Group, Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar, Gujarat (INDIA) 382428
219
- E - Magnets and Power Supplies.
P2T-E-236
HIGH POWER IGBT BRIGE APPLICATION FOR THE HARMONIC
SUPPRESSION IN THE POWER SUPPLY SYSTEM OF THE SPANISH
STELLARATOR TJ-II.
KIRPITCHEV IGOR, P. MENDEZ1, M. BLAUMOSER1, M.VISIERS2, A. AGUDO2, J. IGLESIAS3
1. Asociación EURATOM-CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid, Spain,
2.ENERTRON, S.A. C/Amsterdam Pol. Ind. Torres de Alameda, 28813 Madrid, Spain, 3.CEDEX Alfonso XII, 3 y 5, 28014
Madrid, Spain
The magnetic system of TJ-II is equipped with one toroidal and six poloidal mutually coupled coils. The
central poloidal coils system consists of a solenoid and two helical coils which spiral around this solenoid.
These three coils are placed inside the toroidal system very close to the vacuum vessel and consequently to the
plasma (about 10cm), and that's why the power supply system have to guarantee very low current perturbation
in these coils in order to avoid negative influence on the plasma confinement. The coils systems are supplied
separately by 12 pulse controlled thyristor converters with the maximum DC current 12 kA for the poloidal
system and 32 kA for the toroidal one. The thyristor rectifiers are fed by a fly-wheel generator. Its output
voltage is 15kV and its output frequency varies from 100Hz at the beginning of the experimental pulse to
80Hz at his end. The thyristor rectifiers can vary the current in the coils from zero to their maximum values
and theirs phase control regulators produce a high frequency harmonics in the DC current. The oscillations of
the various regulation systems operating simultaneously and small asymmetries of generator and transformers
produce low frequency sub-harmonics also. The current ripple requirements have been specified to be kept to a
very low level, 1% of the actual coil current in all coil systems but not more then +25 A in the poloidal coils
and not more than +50 A in the toroidal coil. Six years of TJ-II operation demonstrate that current ripple
requirements are being met, but nevertheless a further reduction of about one order of magnitude namely to
+2A is necessary. The investigations of different methods show that the active filter is the most appropriate
way for the current ripple reduction. Different computer simulations have been carried out in order to confirm
the feasibility of the technical solution and to define the main parameters of the filter. The results of the
calculations demonstrate that the active filter connected in parallel to the load and based on IGBT H-bridge
can reduce the current ripple of the coil to the specified +2A limit. Spanish company ENERTRON has
manufactured the active filter and the tests at factory have confirmed the correct operation of the equipment.
The paper describes in detail the design of the filter, the computer simulations, the results of a test circuit at
factory and the results of the commissioning tests on site.
Corresponding Author:
KIRPITCHEV IGOR
igor.kirpitchev@ciemat.es
Asociación EURATOM-CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid,
Spain
220
- E - Magnets and Power Supplies.
P2T-E-259
MANUFACTURE AND TEST OF THE NON-PLANAR COILS FOR
WENDELSTEIN 7-X
RUMMEL, THOMAS, KONRAD RISSE HARTMUT EHMLER
Max-Planck-Institut für Plasmaphysik, Euratom Association, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald,
Germany
The standard magnetic configuration of WENDELSTEIN 7-X (W7-X) is formed by 50 non-planar
superconducting coils. 20 additional planar superconducting coils allow to modify the magnetic configuration.
Due to the symmetric arrangement of 5 equal modules, each being composed of two mirror-symmetric halfmodules, 5 differently shaped types of non-planar coils are sufficient. The nominal current is 17.6 kA for all
non planar coils, which can be varied between 14.5 kA for 2.5 Tesla operation up to 18.3 kA for the 3 Tesla
operation in the low shear scenario. The winding of the coils is made of 108 turns of a forced flow cable-inconduit conductor using a NbTi superconductor. It consists of a rope with 243 strands enclosed in an
aluminium jacket with a void fraction of 37 %. The outer dimensions of the jacket are 16 mm x 16 mm. One
strand has a diameter of 0.57 mm and is made of 144 NbTi filaments stabilized by copper with a copper to
non-copper ratio of 2.7. The specified critical current of the superconductor is 32 kA at 4.2 K and 6 Tesla. To
withstand the electromagnetic forces of up to 400 t each winding pack is stiffened by a massive steel casing,
which leads to a total weight of a non-planar coil of about 5.5 tons. Typical dimensions of a non-planar coil are
about 3.5 m x 2.5 m x 1.5 m. The contribution gives a report about the status of the production comprising the
production of the winding packs and the casings as well as the assembly of the coil, including instrumentation.
Special attention had to be given to the quench detection wiring. The design of the quench detection cables,
which consist of 6 single wires each, was changed in order to ensure a better electrical strength also under
vacuum conditions. Several coils are already finished and tested at room temperature at the manufacturer’s
site. The test procedure and typical test results will be presented. After production all coils will be tested under
cryogenic conditions, too. Main tests are a nominal current test, a quench test by increasing the temperature, a
high voltage test, a helium leak test and measurement of the stresses in the casing, the shrinkage during cool
down and the deformations due to electromagnetic forces. The test procedure and the results of the first tests of
the coils will be presented and compared with the expectations.
Corresponding Author:
RUMMEL, THOMAS
thomas.rummel@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Euratom Association, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491
Greifswald, Germany
221
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P2T-E-266
V-I CHARACTERISTICS WITH BUMPS IN THE MEDIUM SIZE NBTI
CICC CABLES.
BRUZZONE PIERLUIGI, BORIS STEPANOV (1) ELENA ZAPRETILINA (2)
(1) EPFL-CRPP, CH-5232 Villigen-PSI, Switzerland (2) NIIEFA (D.V.Efremov Institute), 186641, St. Petersburg, Russian
Federation
In the scope of conductor R&D program for the poloidal field coils of the ITER fusion project, a number of
short samples of sub-size NbTi cable-in-conduit conductors with 336 strands (CICC) and full size (1440
strands) have been tested in the SULTAN facility (CRPP Villigen, Switzerland). The cabling and jacketing
work is done at VNIIKP (Moscow). The dc test (critical current and current sharing temperature) was carried
out with supercritical helium at 10 bar, over a broad range of operating temperature (4.5 – 7 K) and
background field (4 – 7 T). The paper reports about two different cases of “bumpy behaviour observed in the
voltage-current characteristics. During the tests at increased operating temperature, an abnormal behaviour of
the voltage – current characteristic has been observed. In some cases, instead of smooth, monotonous
transition (power law V-I characteristic) a wave-like voltage development and recovery (“bumps”) have been
seen. The most likely reason for this abnormality seems to be minor temperature variations, in the range of few
hundredths of degree, caused by minor pressure waves in the supercritical helium circuit. The paper discusses
conditions at which the effect could be observed, and presents some analysis which reproduces the observed
behaviour and support the hypotheses about ‘thermal’ origin of the voltage ‘bumps’. Another, different
“bumpy” behaviour in the current-voltage characteristic is reported, where the reason is discussed to be linked
with a current re-distribution phenomenon, also supported by the response of Hall sensors monitoring the selffield of the sample.
Corresponding Author:
BRUZZONE PIERLUIGI
pierluigi.bruzzone@psi.ch
EPFL-CRPP, CH-5232 Villigen-PSI, Switzerland
222
- E - Magnets and Power Supplies.
P2T-E-284
HIGH TEMPERATURE SUPERCONDUCTORS FOR THE ITER
MAGNET SYSTEM AND BEYOND
FIETZ, WALTER H.(1), STEFAN FINK(1) REINHARD HELLER(1) PETER KOMAREK(1) VIPULKUMAR L.
TANNA(1) GERNOT ZAHN(1) GABRIEL PASZTOR(2) RAINER WESCHE(2) ETTORE SALPIETRO(3)
ALEXANDER VOSTNER(3)
(1) Forschungszentrum Karlsruhe, Institut für Technische Physik, Karlsruhe, Germany. (2)Centre de Recherches en Physique
des Plasmas, Villigen, Switzerland. (3) European Fusion Development Agreement, Close Support Unit, GARCHING,
Germany.
Selected also for Oral Presentation
O2A-E-284
Operation currents up to 68 kA have to be transferred from room temperature (RT) to 4.5 K for the
superconducting magnet system of ITER. This current transfer is made using specially designed current leads
(CL). With the conventional design of such CL, ohmic losses cause high heat loads to the refrigeration system
which is critical in the low temperature range where the efficiency of the refrigerator is reduced according to
the Carnot rule. For ITER a cooling power of 64 kW at 4.4 K is foreseen taking more than 20 MW of electric
power. This large power consumption can be reduced drastically by the use of High Temperature
Superconductor (HTS) current leads for ITER and future fusion machines because these HTS CL have no
ohmic losses in the range of 4.5 K to 70 K. In the frame of the European Fusion Technology Programme, the
Forschungszentrum Karlsruhe and the CRPP Villigen have designed and built a 70 kA current lead using HTS
material. This HTS current lead was installed and tested in the TOSKA facility of the Forschungszentrum
Karlsruhe. The experiment covered the electrical and thermal behaviour under steady state conditions and in
case of a quench. To characterize the performance of the current lead, the temperature profile, the contact
resistances, the required cooling power, and the critical current were evaluated. To check extreme conditions a
complete loss of He-flow was studied, too. Results of the experiments carried out in TOSKA facility are
presented. In addition an outlook of future prospects of HTS material applications in a fusion machine will be
given. An obvious possibility is to introduce HTS in the RT bus bar system to reduce losses allowing a much
lower effort for thermal shielding and possibly alternative thermal insulation concepts. A preliminary layout of
a HTS bus bar system is shown. However, the real challenge is to use HTS materials for the whole magnet
system. Even when such a magnet system is operated at 20 K, those fusion machines would be much more
efficient due to the reduction of electric power consumption for cryogenics by a factor of 5-10. An improved
version would be a machine with a magnet system operating at 65 K to 77 K, because in this case liquid
nitrogen could be used as coolant. An overview about status, promises and challenges of HTS conductors on
the way to an HTS fusion magnet system beyond ITER is given.
Corresponding Author:
FIETZ, WALTER H.(1)
Walter.Fietz@itp.fzk.de
Forschungszentrum Karlsruhe, Institut für Technische Physik, Hermann-von-Helmholtz-Platz 1, D-76344
Eggenstein-Leopoldshafen
223
- E - Magnets and Power Supplies.
P2T-E-299
ANALYSIS OF THE RESISTIVE TRANSITION IN NB-TI CABLE-INCONDUIT CONDUCTORS VIA AN EXTENDED 1-D MODEL
ZAMBONI WALTER, PIERLUIGI BRUZZONE (1), LUCA BOTTURA (2), CLAUDIO MARINUCCI (1)
(1) CRPP-Technologie de la Fusion, CH-5232 Villigen-PSI, Switzerland (2) CERN, AT-MTM, CH-1211 Geneva 23,
Switzerland
In the frame of the research on applied superconductivity, the resistive transition of superconducting cables is
one of the most discussed issues in the recent years. Experimental tests performed on NbTi middle-size cablein-conduit conductors (CICCs) at CRPP show that the resistive transition parameters, i.e. the critical current
and the n-index, are strongly affected by the self (magnetic) field effect and current redistribution. The strong
self-field gradients, due to large operating current, induce a current redistribution among strands. The whole
phenomenon is strongly dependent on the values of transverse resistivity of the cable, which is able to either
avoid or promote current redistribution. In order to investigate the influence of the self field, we simulate the
current sharing process in a double stage model of the cable, which takes into account longitudinal and
transverse resistive effects. To this purpose, we retain a power-law model for the basic superconducting
element. In the evaluation of the electromagnetic behavior of the cable, we adopt an extended 1-D approach. It
consists of a Multiconductor Transmission Line (MTL) Model with constant inductive and resistive transverse
coupling between elements of the MTL. The magnetic model is coupled to an assessed thermohydraulic one
and numerically solved by CryoSoft Thea®. We tested our model against the critical current experiments
performed on different conductors, which mainly differ in the interstrand resistance. The results show that the
model, although not complete and “self-consistent” as a 3-D one is, can be a reliable and relatively fast tool in
the investigations on the self-field and current redistribution effects on CICCs, once the strands and the basic
components have been characterized.
Corresponding Author:
ZAMBONI WALTER
zamboni@unina.it
Association EURATOM-ENEA-CREATE, DIEL, Università degli Studi di Napoli "Federico II", via Claudio 21,
80125 Napoli Italy
224
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P2T-E-304
EFFECTIVE BENDING STRAIN ESTIMATED FROM IC TEST RESULT
OF D SHAPED NB3AL CICC COIL FABRICATED WITH A REACT AND
WIND PROCESS FOR THE NATIONAL CENTRALIZED TOKAMAK
ANDO TOSHINARI, KIZU KANAME(1) MIURA YUUSHI(1) TSUCHIYA KATSUHIKO(1) MATSUKAWA
MAKOTO(1) TAMAI HIROSHI(1) ISHIDA SHINICHI(1) KOIZUMI NORIKIYO(1) OKUNO KIYOSHI(1)
(1)Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukouyama, Naka-machi, Nakagun, Ibaraki-ken 311-0193, Japan
In Japan the National Centralized Tokamak is being planned, based on a modification of JT-60 with
superconducting coil. The design of the TF coil is characterized by the maximum magnetic field of 7.4 T at a
nominal operating current of 19.4 kA and by the use of a react and wind process with the maximum bending
strain of 0.4 % on the Nb3Al cable using a Nb3Al CIC conductor. The bending strain is defined from
theNb3Al cable diameter divided by the winding diameter. Nb3Al is insensitive to strain on Ic in comparison
with Nb3Sn. The conductor consists of 216 Nb3Al strands and 108 copper wires inserted into the circular hole
of a rectangular stainless steel conduit. In order to demonstrate the applicability of Nb3Al conductor with the
react and wind process to the TF coil, a two turns-D shaped Nb3Al coil whose height is about 2 m, with the
full size CIC conductor has been fabricated and tested by installing its corner part wound with a bending strain
of 0.4 %, into a split coil as background field. In this test the strain corresponding to the degradation of Ic on
the Nb3Al conductor due to its bending, so called the effective bending strain to be converted into the axial
strain, was investigated and estimated. From Ic results in this test the total strain on the conductor was
estimated as – 0.57 %. The estimation was carried out taking account of the magnetic field and strain
distribution within the Nb3Al conductor. On the other hand, the axial strain due to the thermal stress from the
stainless steel conduit on the Nb3Al filament was found to be -0.57 % from the other experiment. This means
that the degradation of Ic due to the bending is neglect. Therefore, the effective bending strain on the Nb3Al
conductor was zero. It is considered that the strands in the cable slipped each other toward the reduction of
bending strain during the bending. This result is very useful for the fusion coil fabrication with the react and
winding process. In this paper, the effect of bending strain on Ic in Nb3Al cable-in-conduit conductor is
discussed.
Corresponding Author:
ANDO TOSHINARI
ando@naka.jaeri.go.jp
Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801-1 Mukouyama, Nakamachi, Naka-gun, Ibaraki-ken 311-0193, Japan
225
- E - Magnets and Power Supplies.
P2T-E-306
ELIMINATION OF VARIABLE HARMONICS ON MOTOR
GENERATOR CIRCUIT FOR EXPERIMENTAL FUSION FACILITY
YAMADA SHUICHI, NAKANISHI YOSUKE (1) KOJIMA HIROSHI (1) HIUE HISAAKI (2) UEDE TOSHIO (2)
MITO TOSHIYUKI
(1) Fuji Electric Advanced Technology Co. Ltd., Fuji-machi, Hino, Tokyo 191-8502, Japan. (2) Fuji Electric Systems Co.
Ltd., 1-1, Tanabeshinden, Kawasaki, 210-9530, Japan
In an experimental fusion device, a large electric power is needed for producing high temperature plasma in
the high density regime. Since the motor generator with a flywheel (FW-MG) can generate large electric
power without giving the turbulence to the commercial power grid, it is used for the back power source of the
heating devices such as the NBI, ECH and/or ICRH. The frequency of FW-MG changes almost factor of two
between starting phase and running down phase during a pulse. When the power supply of the heating device
is composed of full wave rectifiers using the thyristors, the harmonic currents of the 5th, 7th and other higher
components appear on the output circuit of the FW-MG. The frequencies of these harmonic currents also
change the same in proportion to the fundamental frequency of the FW-MG. These variable harmonic currents
may threaten to damage the windings of the generator and/or the transformers of heating devices caused by the
abnormal temperature increase. To avoid these deteriorations, a special active filter, which can eliminate the
variable harmonic currents in continuity, was investigated. It has the following major functions; 1) the
detection of the variable frequency of the power line, 2) the extraction of current component of fundamental
frequency, 3) the operation of current component of higher harmonics, and 4) the compensation of harmonic
current by generating the counter-flow current. A special algorithm of the band-pass filter was developed for
the extraction of current component of fundamental frequency. Dynamic simulations for the active filter, FWMG and power supplies of heating devices for the experimental fusion device of LHD has been conducted by
using the analysis tool of the PSCAD/EMTDC. We confirmed that the harmonic currents with the amplitudes
of 20% were suppressed to less than 2 % through the operational frequency range from 95 Hz to 55 Hz by
using this active filter.
Corresponding Author:
YAMADA SHUICHI
yamadas@LHD.nifs.ac.jp
National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, Japan
226
- E - Magnets and Power Supplies.
P2T-E-311
FATIGUE ASSESSMENT OF THE ITER TF COIL CASE BASED ON JJ1
FATIGUE TESTS
HAMADA KAZUYA, NAKAJIMA HIDEO, KATSUTOSHI TAKANO AND OKUNO KIYOSHI
801-1, Muko-yama, Naka-machi, Naka-gun, Ibaraki, Japan
In the International Thermonuclear Experimental Reactor (ITER), a structure material for Toroidal Field (TF)
coil case at the inboard leg requires a high strength (0.2% yield strength>1000 MPa) and toughness (fracture
toughness KIC(J) >200 MPam0.5) at 4.5K. Japan Atomic Energy Research Institute (JAERI) has developed
JJ1 (0.05C-12Cr-12Ni-10Mn-5Mo-0.2N) for this application. Since 60,000 cycles of electromagnetic load will
be loaded on the ITER TF Coil case during coil life, a fatigue characteristic of TF coil is important subject for
structural design. There are two approaches for evaluation of fatigue life assessment. One is a fatigue crack
growth rate (FCGR) evaluation, which has been performed based on the measurement of the FCGR. The other
is a fatigue life assessment based on Stress-fatigue life(S-N curve), which we think more appropriate approach
and had not been performed well because measurement data were not enough for establishment of S-N curve
at 4.5K. JAERI has measured the fatigue life of the base metal and welded joint of JJ1 at 4.5K, based on JIS
Z2283 ‘Method of low cycle fatigue testing for metallic material in liquid helium.’ The fatigue test has been
performed in a fully reversed axial - strain controlled method. The strain range and number of cycle are -0.6%
to +0.6% and 10,000 to 2,000,000 cycles, respectively. Total 17 samples have been tested. As a result, failure
cycle at welded joint is evaluated to be 60,000 at stress amplitude of around 740 MPa from the S-N curve
established. The S-N curve of JJ1, together with the safety factor of 20 for failure cycle and 2 for stress
amplitude indicates that the stress amplitude of TF coil case should be kept less than 370 MPa to achieve
required operation cycle of 60,000 in case of JJ1. In TF coil case inboard leg, the severest cyclic stress occurs
at poloidal shear key region located at the top and bottom corners. The recent stress analysis indicates that
maximum principal stress is 600 MPa and cyclic stress is 368 MPa. This stress condition corresponds to the
equivalent alternating stress of 250 MPa, using well-known Goodman’s diagram method, which is below the
allowable value. Therefore, it is concluded that the JJ1 satisfied the specified operation cycle of 60,000.
Corresponding Author:
HAMADA KAZUYA
hamada@naka.jaeri.go.jp
801-1, Muko-yama, Naka-machi, Naka-gun, Ibaraki, Japan
227
- E - Magnets and Power Supplies.
P2T-E-326
EFFECT OF ELECTRICAL CHARACTERISTICS OF SIC POWER
DEVICE ON OPERATIONAL EFFICIENCY OF AC/DC CONVERTER
TATSUYA MATSUKAWA, HIROTAKA CHIKARAISHI (1) YOSHIHISA SATO (2) RYUICHI SHIMADA (3)
(1)National Institute for Fusion Science, Oroshi-cho, Toki, Gifu, JAPAN (2)Daido Institute of Technology, Takiharu-cho,
Minami-ku, Nagoya, JAPAN (3)Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Ookayama,
Meguro, Tokyo, JAPAN
The large capacity AC/DC converter, which is to output high DC current for energizing the magnetic field coil
of nuclear fusion experimental facility or SMES system, is designed to minimize its operational loss for high
efficiency operation. SiC power device is a new power electronics device based on crystallized SiC material,
and is one of the most promising switching elements to be applied to large capacity AC/DC converter. SiC
power device is expected to have more excellent electrical characteristics than those of conventional Si power
device. Typical advantages of SiC power device in electrical characteristics are high withstanding voltage, low
on-state resistance, high operational temperature, fast switching speed, etc. Concerning the efficiency
improvement of AC/DC converter, the temperature dependence of electrical characteristics of switching
element is the important issue on operational loss reduction. On-state resistance and allowable operational
temperature of switching element are directly related to the operational loss of AC/DC converter. The former
one is the main parameter of conductive loss, which is dominant one of the operational loss of high current
AC/DC converter, and it is effective to reduce the conductive loss for efficiency improvement. The latter one
is related to the cooling capability of cooling equipment, which is a main auxiliary component of large
capacity AC/DC converter. The temperature dependence of on-state resistance of SiC power device will be
expected to reduce the conductive loss and to simplify the cooling equipment of AC/DC converter. The
allowable operational temperature of SiC power device is higher than that of Si power device, therefore it
allows also to minimize the cooling capability. Both electrical characteristics of SiC power device will
contribute to reduce the operational loss and to improve the operational efficiency of AC/DC converter.
Previously, the operational loss of large capacity AC/DC converters of ITER class power supply was studied
based on the predicted on-state resistance of future SiC power device. In this paper, with the results of
measurement in some experiments for present unipolar SiC device, the effect of electrical characteristics of
SiC power device is mainly discussed on efficiency improvement. And, the temperature dependence of the
electrical characteristics will be mentioned in comparison with conventional Si power device.
Corresponding Author:
TATSUYA MATSUKAWA
matukawa@elec.mie-u.ac.jp
Department of Electrical and Electronic Engineering, Mie University, 1515, Kamihama-cho, Tsu, Mie, 514-8507,
JAPAN
228
- E - Magnets and Power Supplies.
P2T-E-338
DESIGN REQUIREMENT, QUALIFICATION TESTS AND
INTEGRATION OF A THIN SOLID LUBRICANT FILM OF MOS2 FOR
COLD MASS SUPPORT STRUCTURE OF THE STEADY STATE
SUPERCONDUCTING TOKAMAK SST-1.
DOSHI BHARAT1, B. SARKAR1, PRATIMA REWATKAR1, DASHARATH SONARA1, C.RAMDAS1, K.J.THOMAS1,
SAXENA Y.C.1, V.R.BHASKAR2, SENTHIL KUMAR2 AND S.NAGBHUSHANAM2
(1) Institute for Plasma Research, Bhat, Gandhinagar, Gujarat, India 382 428 (2) ISRO Satellite Center,
Bangalore, India The SST-1 is a super conducting tokamak, which is in the final phase of assembly and
commissioning. The super conducting magnet system of SST1 comprises of Toroidal field (TF) and Poloidal
field (PF) coils. The 16 TF coils are nosed and clamped towards the in-board side and are supported toroidally
with inter-coil structure at the out-board side, forming a rigid body system. The 9 PF coils are clamped on the
TF coils structure. The integrated system of TF coils & PF coils forms the cold mass of @ 50 Ton weight. This
cold mass is accommodated inside the cryostat and freely supported on the rigid support ring at 16 locations
and support ring in-turn supported on 8 columns of machine support structure. During the operation this cold
mass attains a cryogenic temperature of 4.2K in the hostile environment of high vacuum. The thermal
excursion of cold mass and its supporting structure during this cool down results into severe frictional forces at
the supporting surfaces. There is a design requirement of introducing a thin layer of solid lubricant film of
MOS2 having coefficient of friction 0.05 between the sliding surfaces to control the stress contribution due to
the friction. To ascertain the compatibility of molybdenum disulphide (MOS2) as a solid lubricant in high
vacuum and very low temperature environment, we have carried out qualification tests on various samples and
measured the coefficient of friction in both the room temperature conditions and at high vacuum & after
thermal shocking to 4.2K temperatures. After successful qualification tests actual components are fabricated
and integrated in the cold mass support structure assembly. This paper presents the design requirement,
qualification tests performed and details about the integration of thin solid lubricant film of MOS2.
Corresponding Author:
DOSHI BHARAT1
doshi@ipr.res.in
Institute for Plasma Research, Near Indira Bridge, Bhat, Gnadhinagar 382 428 INDIA
229
- E - Magnets and Power Supplies.
P2T-E-344
HOW SHOULD WE TEST THE ITER TF COILS ?
LIBEYRE PAUL, CIAZYNSKI DANIEL DUCHATEAU JEAN-LUC SCHILD THIERRY (1) FIETZ WALTER
H. (2) ZAHN GERNOT (2)
(1) CEA/DSM/DAPNIA CEA Saclay F-91191 Gif-sur-Yvette cedex (France) (2) Association Euratom-Forschungszentrum
Karlsruhe, Institut für Teschnische Physik, Helmholtzplatz 1 D-76344 Eggenstein-Leopoldshafen (Germany)
The ITER TF coils are a major piece of the ITER magnetic system. It is of prime importance that they operate
reliably during the whole life of the machine, since a failure in these coils during operation would cause a
major breakdown in the programme and would lead to a difficult repair procedure. The manufacture of these
coils will thus include a very strict quality assurance system at each step from the strand production to the final
closure welding of the case around the winding-pack in order to avoid any defect. Nevertheless, these coils
will be the largest Nb3Sn coils ever built and will operate at high magnetic field with high current and will be
submitted as well to very large mechanical loads as high voltages. It is therefore necessary that their behaviour
be well established. The paper addresses the advantages and drawbacks of the main options which can be
considered. Particular attention is put on the information provided by testing at helium temperature each
completed coil with respect to extensive testing of component samples. The critical points to be checked are
the dielectric strength of the insulation, the internal joints resistance, the temperature margin of the conductor,
the hydraulic resistance of the cooling circuit. Several testing configuration are considered and their impact in
terms of cost and time schedule are estimated. An other important aspect of the cold tests is to verify the
correct operation of the coil safety system in relevant conditions. The cold tests are the final reception tests of
the coils and should therefore bring the guaranty that the coil will be able to operate safely at its nominal field
and current. When assembled together to form a toroid, the operating conditions of the TF coils are highly
depending on the other coils of the magnet. The maximum field applied to the conductor is reaching 11.8 T at
a nominal current of 68 kA when the coil is inside the TF magnet, whereas when tested alone, the maximum
field is hardly exceeding 6 T. On the other hand, in a single coil test the coil experiences only in-plane loading,
whereas during plasma operation both in-plane and out-of-plane loads are applied to the coil. In the purpose of
achieving more relevant operating conditions, an investigation of the possibility of testing simultaneously
several coils is performed and discussed.
Corresponding Author:
LIBEYRE PAUL
libeyre@cea.fr
Association Euratom-CEA CEA/DSM/DRFC CEA Cadarache F-13108 St Paul lez Durance cedex (France)
230
- E - Magnets and Power Supplies.
P2T-E-379
CYCLIC TESTING OF SHEAR KEYS FOR THE ITER MAGNET
SYSTEM
ROSSI PAOLO, L.F. MORESCHI (ENEA CR BRASIMONE) A. PIZZUTO (ENEA CR FRASCATI) S. STORAI
(ENEA CR BRASIMONE) C. SBORCHIA (EFDA CSU)
ENEA CR Frascati, PB 65, 00044 Frascati, (Roma), Italy ENEA CR Brasimone, PB 1, 40032 Camugnano BO), Italy EFDA
Close Support Unit, Boltzmannstrasse 2, D-85748 Garching, Germany
Shear keys are to be used to support the out-of-plane loading of the Toroidal Field (TF) coils during a plasma
pulse in ITER. At the Inner Intercoil Structures (IIS) a set of poloidal shear keys is used to take the shear load
at each connection between adjacent TF coils. Solid circular keys have been selected as reference. At the Outer
Intercoil Structures (OIS) adjustable conical shear keys and friction joint based shear panels are used to take
the shear load. Low voltage electrical insulation is required at the flanges of the IIS and OIS, plus for all the
bolts, poloidal keys and adjustable keys. This electrical insulation has to withstand large compression
associated with some shear or slippage. A ceramic coating was selected for this purpose. The main scope of
the experimental campaign was the mechanical testing of the shear keys and the electrical insulation in
operational conditions relevant to ITER. Both keys were made of Inconel 718, provided with a ceramic
alumina coating and inserted into flanges made of cast AISI 316 LN. The adjustable conical shear key was preloaded at room temperature and subject to cyclic shear loads of 2.5 MN for a large number of cycles (about
30,000) at cryogenic temperature (77 K). The conical key and the alumina coating resulted undamaged after
test. Another test campaign was then performed with higher shear loads (up to 3 MN) to reach a sufficient
safety margin even with the friction effect due to the pre-load. A set of 15,000 cycles were completed followed
by some cycles at higher loads to reach the ultimate limit, which is the shear load to be experienced by the key
in case of a Poloidal Field (PF) coil short.
Corresponding Author:
ROSSI PAOLO
paolo.rossi@frascati.enea.it
ENEA CR Frascati, PB 65, 00044 Frascati, (Roma), Italy
231
- E - Magnets and Power Supplies.
P2T-E-390
MODULAR COIL DESIGN DEVELOPMENTS FOR THE NATIONAL
COMPACT STELLARATOR EXPERIMENT (NCSX)
WILLIAMSON, DAVID E., A. BROOKS (1), T. BROWN (1), J. CHRZANOWSKI (1), M. COLE (2), H-M. FAN
(1), K. FREUDENBERG (3), P. FOGARTY (2), T. HARGROVE (4), P. HEITZENROEDER (1), G. LOVETT (5), P.
MILLER (5), R.L. MYATT (6), B. NELSON (2), W. REIERSEN (1), D. STRICKLER (2)
(1) PPPL, PO Box 451, Princeton, NJ 08540 (2) ORNL, PO Box 2008, Oak Ridge, TN 37831 (3) BWXT, PO Box 2009, Oak
Ridge, TN 37831 (4) Hargrove Engr, Scottsboro, AL 35768 (5) MK Tech, Knoxville, TN 37930 (6) Myatt Consulting, Norfolk
MA 02056
The National Compact Stellarator Experiment (NCSX) is a quasi-axisymmetric facility that combines the high
beta and good confinement features of an advanced tokamak with the low current, disruption-free
characteristics of a stellarator. The experiment is based on a three field period plasma configuration with an
average major radius of 1.4-m, a minor radius of 0.32-m, and a toroidal magnetic field on axis of up to 2-T.
The modular coils are one set in a complex assembly of four coil systems that surround the highly shaped
plasma. There are six each of three coil types in the assembly for a total of 18 modular coils. The coils are
constructed by winding flexible, copper conductor onto a stainless steel winding form that has been cast and
machined to high accuracy, so that the current center of the winding pack is within +/-1.5-mm of theoretical.
The modular coils operate at 80-K and produce the primary magnetic shaping field for a flat-top pulse length
of 0.5-s at a current of 820-kAT. Due to geometry constraints, the coil windings must operate at high current
density and are subject to rapid heating and thermal stress during a pulse. In addition, the coils experience
electromagnetic forces of up to 1.2-MN/m. This paper will discuss the progression of the coil design from
physics targets to filamentary models to prototype components, which are currently being fabricated.
Advances in inspection technology, field error compensation analysis, and assembly simulation will be
highlighted.
Corresponding Author:
WILLIAMSON, DAVID E.
williamsonde@ornl.gov
Oak Ridge National Laboratory, Post Office Box 2008, Oak Ridge, TN 37831-6169
232
- E - Magnets and Power Supplies.
P2T-E-394
CONCEPTUAL DESIGN OF SPHERICAL TORUS WITH TF-CS HYBRID
COILS BASED ON VIRIAL THEOREM
TSUTSUI HIROAKI, NOMURA SHINICHI, TSUJI-IIO SHUNJI, SHIMADA RYUICHI
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo 152-8550, Japan
A conceptual design of a spherical torus (ST) device with a new type of toroidal field (TF) coils and a central
solenoid (CS) whose stress is reduced to the theoretical limit determined by the virial theorem is proposed. In
the last decade, we had developed a tokamak with force-balanced coils (FBCs) which are multi-pole helical
hybrid coils combining TF coils and a CS coil. The combination reduces the net electromagnetic force in the
direction of major radius by canceling the centering force due to the TF coil current and the hoop force due to
the CS coil current. This excellent feature of FBC and its capability of tokamak operation were investigated
and demonstrated by the first FBC tokamak “Todoroki-I'', while working stress in coils has not yet been
investigated whereas the net electromagnetic force is reduced. Next, we had extended the FBC concept using
the virial theorem which shows that strength of magnetic field is restricted by working stress in the coils and
their supporting structure. High-field coils should accordingly have same averaged principal stresses in all
directions, whereas conventional FBC reduces stress in the toroidal direction only. In that work, we had
obtained the poloidal rotation number of helical coils which satisfied the uniform stress condition, and named
the coil as virial-limit coil (VLC). VLC with a circular cross section of aspect ratio A=2 reduces maximum
stress to 60% compared with that of TF coils. A tokamak discharge in VLC was also demonstrated by the first
VLC tokamak “Todoroki-II”. Recently, we have developed a VLC concept with a non-circular cross section,
and reduce the maximum stress to 30% compared with that of TF coils in a two dimensional analysis.
Moreover, the VLC configuration has a low aspect ratio and a strongly elongated cross section with a
triangularity, and is similar to that of a ST, while a VLC is a helical coil. In this work, we analyze three
dimensional stress distributions, and evaluate operation scenarios in VLC ST.
Corresponding Author:
TSUTSUI HIROAKI
htsutsui@nr.titech.ac.jp
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo 1528550, Japan
233
- E - Magnets and Power Supplies.
P2T-E-405
EMI ON DIAGNOSTICS AND CONTROL CIRCUITS DUE TO
SWITCHING POWER SUPPLIES
GAIO ELENA, R. PIOVAN, V. TOIGO
Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, C.so Stati Uniti 4, 35127 Padova, Italy
In fusion experiments, the use of switching power supplies is becoming more and more frequent. These
equipments are characterized by instantaneous output voltages quickly varying with the switching frequency;
moreover, the output terminals, usually connected to the machine windings, present common mode voltages
varying very fast as well. The typical voltage variation speed is in the order of some kV / microseconds. The
common mode voltages can excite the circulation of noise currents through parasitic capacitances, which
couple different parts of the circuit designed to be isolated one to each other. The harmonic spectra of these
currents range between some hundreds of kHz to some MHz. The windings are usually connected to ground
through high impedances; nevertheless, the parasitic capacitances between the coils, the mechanical supporting
structure and ground represent low impedance electric connections in high frequency ranges. As a
consequence, currents can flow in the reference potential grids, which disturb the plasma diagnostic
equipments; moreover, direct capacitive coupling between the switching power supply loads and the
diagnostics themselves can cause the circulation of noise currents in the measurements circuits. These general
problems have been analyzed in detail for the RFX case, where switching power supply have been introduced
both in the poloidal and toroidal circuits for performing various control actions on the plasma. The analyses
showed that without any provisions the common mode currents in the machine reference potential conductors
can reach values in the order of amperes. In RFX, EM interferences, again referable to the same phenomenon,
were observed in the control sections of the power supplies too. Faults on the drivers of dc static current
breakers happened, due to the presence of very high noise currents in the isolated power supplies of the
switching devices’s firing system; the currents were produced by the fast varying common mode voltage
present at the semiconductors power terminals. Equivalent electric schematizations of these phenomena have
been derived and numerical simulations have been worked out, which explain the induced noises and have
been utilized to identify suitable correction measures; in the paper these analyses are reported and the
experience developed in coping with the reduction of these types of EMI interferences is described.
Corresponding Author:
GAIO ELENA
elena.gaio@igi.cnr.it
Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, C.so Stati Uniti 4, 35127 Padova, Italy
234
- E - Magnets and Power Supplies.
P2T-E-406
THE CONTROL SYSTEM OF THE TOROIDAL POWER SUPPLY OF
RFX
PIOVAN ROBERTO, V. TOIGO (1), L. ZANOTTO (1), M. PERNA (2), A. COFFETTI (2), M. FREGHIERI(2), M.
POVOLERO (2)
(1)Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 (2)ASIRobicon, Viale Sarca
336 - 20126 Milano, Italy
This paper deals with the control system of the rebuilt toroidal field power supply of RFX. In a Reversed Field
Pinch, such as RFX, the waveform of the toroidal field is much more complicated than in a tokamak, as the
current in the toroidal field winding needs to be reversed in polarity and then regulated to produce either
rotating m=0 harmonics or Poloidal Current Drives. The toroidal field power supply of RFX, which represents
the first example of static power supply for fusion experiment based on components named IGCTs, combines a
very complex circuit topology and a fast, reliable and flexible control system. The peculiarity of this system is
the integration of many control functions in a compact solution; the control design has required particular
efforts due to the different kind and nature of devices working together, which includes large capacitor banks
and fast and unconventional power electronics apparatus, such as dc/ ac inverters, choppers and static circuit
breakers. Both slow and fast control algorithms regarding the supervision of the plant, the protection system
and the current and voltage regulation in the winding sectors are implemented in a unique hardware
arrangement. The design criteria of the toroidal power supply control system will be discussed in the paper.
The most important guideline in defining the specifications has been the flexibility, which is strictly related to
the possibility of easily changing the experimental set-up according to requests coming from different
scenarios. Moreover, to simplify the commissioning of the power supply, efforts have been put in designing a
special part of the control section to perform local tests. In such a way an easy setting-up of the system
parameters is possible, thus allowing to test both a single device and the whole system during an experimental
sequence. Another important issue discussed in the paper is the definition of the active plant protection
strategies: the complexity of the system, which presents many different fault cases and operative scenarios, has
led to very fast and sophisticated protection algorithms, integrated in the control structure. The control
hardware architecture derived from the above-mentioned considerations is based on two VME crates, each
including a PowerPC board and some FPGA general-purpose boards. The paper will present a detailed
description of the control functions and of the hardware structure; the software architecture will be also
described.
Corresponding Author:
PIOVAN ROBERTO
roberto.piovan@igi.cnr.it
Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127
235
- E - Magnets and Power Supplies.
P2T-E-417
COMPONENTS AND SYSTEM TESTS ON THE RFX TOROIDAL
POWER SUPPLY
PERNA MAURO, V. TOIGO (1), L. ZANOTTO (1), E. GAIO (1), P. BORDIGNON (2), A. COFFETTI (2), R.
NOVARO (2), P. BERTOLOTTO (3), E. RINALDI (3), G. VILLA (3)
(1) Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, 35127 Padova, Italy (2) ASIRobicon,
Viale Sarca 336 - 20126 Milano, Italy (3) Passoni&Villa, viale Suzzani 229 - 20162 Milano, Italy
The final Site tests of the new RFX toroidal power supply system have been concluded by the end of 2003 and
the system is now ready for the integration tests with RFX machine. This system represents the first example
of static power supply for fusion experiments, based on recent power semiconductors named IGCTs
(Integrated Gate Commutated Thyristors). Besides producing the time-dependent toroidal magnetic field
required to set up the RFP configuration, it is also devoted to generate and control the toroidal magnetic field
waveforms required to produce rotating m=0 field harmonics for applying a torque to the plasma or special
Poloidal Current Drives. The system design was peculiar not only for the single components, which are
developed “ad hoc” and characterised by a high degree of technological innovation, but also for the system
coordinated operation, which is unique and very articulated, due to the high flexibility level required. Special
factory tests on the main device prototypes have been performed to verify the critical aspects of the design. In
the static breaker (4 kV, 16 kA, 128 MA2s), which represents a remarkable example of static dc current
interruption technology at high power, it was necessary to realize five parallel IGCTs branches. Reaching good
performances in terms of current sharing and limited reapplied overvoltages was not so straightforward. Also
for dc/ac inverters (3 kV, 6 kA), composed of three single-phase IGCTs H-bridges in parallel, current sharing
optimization has been a very ambitious goal: they represent the first industrial realization of parallel
connection among IGCTs bridges; for both, the test results were very satisfactory. In the capacitor bank design
(4 kV, 16 mF, 128 kJ), the most peculiar aspect is the protection against internal faults: the fuse design had to
satisfy many different requirements: very fast intervention, less than 100 ms, fault discrimination at different
bank voltage levels, no explosion in the worst fault conditions; also in this case, special tests have been
performed. To achieve the required coordinated operation of all these devices was also a big task; all the
necessary tests were performed on Site; the control system was designed to assure a high flexibility level and
allowed the integration of one device at a time. The most peculiar aspects of the integration tests and the
optimisation of the whole system operation will be described and discussed in the paper.
Corresponding Author:
PERNA MAURO
mauro.perna@it.asirobicon.com
ASIRobicon, Viale Sarca 336 - 20126 Milano, Italy
236
- E - Magnets and Power Supplies.
P2T-E-420
COMMISSIONING AND OPERATION OF 130KV/130A SWITCHEDMODE HV POWER SUPPLIES WITH THE UPGRADED JET NEUTRAL
BEAM INJECTORS
DAVID C EDWARDS, MARCO BIGI (1), DENIS BROWN (1), DANIEL GANUZA (2), FRANCISCO GARCIA (2),
ZACHARY HUDSON (1), TIMOTHY JONES (1) AND ALBERTO PEREZ (2)
(1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB UK (2) Grupo JEMA, E-20160
Lasarte-Oria, Guipúzcoa, Spain
The JET neutral beam (NB) upgrade increased the current of eight Positive Ion Neutral Injector (PINI) beam
sources from 30A to 60A at 130kV acceleration voltage. This involved: (1) procurement of two new HV
Power Supply (HVPS) units each rated at 130kV/130A (20s pulse length, 1/30 duty cycle) to feed two
130kV/60A PINIs; (2) reconfiguration of four existing HVPS modules (rated at 160kV/60A) to feed four
individual PINIs. The new HVPS units are of all solid-state switched-mode type, where 120 high-frequency
IGBT invertor modules feed 120 isolation transformers whose rectified outputs are connected in series.
Regulation and fast switching for control and load protection are done on the LV side, and the isolation
transformers provide a passive barrier to the transfer of energy in case of IGBT failure. Up to 255 reapplications allow for repetitive HV breakdowns in the load, detected when the current exceeds a pre-defined
first threshold. Two independent optically triggered thyristor crowbars across the HV output act if the load
current reaches a second threshold. The design of the HVPS will be discussed, including load protection
characteristics in comparison to conventional switch-tube designs. The paper describes on-site testing,
commissioning and operation of the new HVPS units. Both units were factory tested at full power and pulse
length into dummy resistive load. Following on-site installation, the factory tests were repeated. The transition
from dummy-load testing to PINI load operation required full integration of the HVPS within the overall JET
control system, and rigorous testing of the co-ordinated actions and protections of all PINI power supplies
(filament and arc for plasma source and negative suppression grid). The implementation of the ‘fast logic’
electronics for these functions will be described. Extensive use was made of parasitic integrated test pulses,
where the other PINIs could be operated normally, and the HVPS was energised under full remote control
together with the corresponding PINI plasma sources, but with the HVPS connected to dummy load (or open
circuit). The amount of NB operation time dedicated to commissioning was thereby minimised, yet gave a
high degree of confidence of readiness for HV energisation of the PINI, and first beam extraction occurred
within less than 24 hours of HV connection to the PINI. Finally, the routine operating experience, such as
performance and reliability, of the new HVPS units will be described.
Corresponding Author:
DAVID C EDWARDS
david.c.edwards@jet.uk
Building J20/1/36, Culham Science Centre, ABINGDON OX14 3DB UK
237
- E - Magnets and Power Supplies.
P2T-E-427
MAGNETIC COMPATIBILITY OF STANDARD COMPONENTS FOR
ELECTRICAL INSTALLATIONS: TESTS ON LOW VOLTAGE CIRCUIT
BREAKERS AND CONTACTORS
GRANDO LUCA, ANTONIO DE LORENZI (1) GIULIO BETTANINI (1) DANIELE DESIDERI (2)
(1) Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova Italy (2) Dept. of Electrical
Engineering, University of Padova, Via Gradenigo 6/a, 35131 Padova Italy
The ITER tokamak building will be permeated by an almost constant magnetic flux density up to 70 mT,
generated by the superconducting winding system. This operation condition asked for a research activity aimed
at assessing the magnetic field immunity of components forming the Low Voltage power supply system. In
this work, the behaviour of some electro-mechanical components and of a Low Voltage cubicle under static or
slowly variable magnetic field has been investigated. A selection of the most recent industrial low voltage
current breakers and contactors of the major European manufactures in the 50 ÷ 250 A range, equipped with
different kind of drivers and protection relays, has been tested by applying a nearly uniform and static
magnetic flux density B up to 80 mT. Off load, on load, protection trip and life tests have been performed for
each component following a specific test procedure and for life tests the directions given by the EN Standards
have been applied. In general, the results showed that the components are sensitive to amplitude, direction and
versus of the magnetic flux density starting from 10 mT. The test campaign has been completed investigating
the behaviour of a fully equipped power center module, consisting in a iron cubicle equipped with different
circuit breakers and contactors (of the same type of those previously tested), remotely controlled by a PLC via
Profibus optical interface –a structure similar to what is expected in the ITER Low Voltage installations – and
immersed in a 40 mT magnetic applied with a repetitive waveform. In order to assess the cubicle shielding
efficiency, a comparison between the magnetic flux density values, computed in vacuum and the magnetic
field measured inside the cubicle has also been performed. Besides, nearly 500 operations have been carried
out switching on and off some resistive loads during the B field pulse. Under these conditions, the overall
behaviour was consistent with the results obtained with the individual components. This work is performed
under the R&D contract 02/1010 between the European Fusion Development Agreement (EFDA) and the
Consorzio RFX.
Corresponding Author:
GRANDO LUCA
luca.grando@igi.cnr.it
Consorzio RFX - Association EURATOM-ENEA, Corso Stati Uniti 4, 35127 Padova Italy
238
- E - Magnets and Power Supplies.
P2T-E-442
FIRST INTEGRATED TEST OF THE SUPERCONDUCTING MAGNET
SYSTEMS FOR THE LEVITATED DIPOLE EXPERIMENT (LDX)
A. ZHUKOVSKY, JOSEPH V. MINERVINI (1) P. C. MICHAEL (1) J. H. SCHULTZ (1) B. A. SMITH (1) J.
KESNER (1) A. RADOVINSKY (1) D. GARNIER (2) M. MAUEL (2)
(1) MIT Plasma Science and Fusion Center 77 Massachusetts Avenue Cambridge, MA 02139 USA (2) Columbia University
Department of Applied Physics and Applied Mathematics Room 210 S. W. Mudd Building New York, NY 10027 USA
The Levitated Dipole Experiment (LDX) is an innovative approach to explore the magnetic confinement of a
fusion plasma offering the possibility of a fusion power source with near-classical energy confinement. In this
concept a magnetic dipole field is created by a superconducting solenoid which is magnetically levitated for up
to 8 hours in the center of a 5-meter diameter vacuum vessel. The Floating Coil (F-coil) is designed for a
maximum field of 5.3 T. A react-and-wind Nb3Sn conductor was selected to enable continuous field
generation as the coil warms from an initial temperature of about 5 K at the start of the experimental day up to
a final temperature of about 10 K at the end of the operating day. The F-coil is maintained in the center of the
plasma volume by the Levitation Coil (L-coil). This coil is made from high temperature superconductor (Bi2223) to minimize the electrical and cooling power needed for levitation. It is a 2800 turn, double pancake
winding that supports the weight of the F-coil and controls its vertical position within the vacuum chamber.
Since the F-coil must operate in a levitated position it is not desirable to have electric or cryogenic feeders
serving the coil through the plasma. For this reason, the coil is inductively charged/discharged and cooled
cryogenically in a lower charging station. The operating current in the F-coil is induced by the Charging Coil
(C-coil) when it is resting in the charging port at the bottom of the LDX vacuum vessel. The C-Coil is a
superconducting solenoid using NbTi operating in a liquid helium cryostat that surrounds the 1157 mm
diameter charging station. It stores 8 MJ of energy at an operating peak field in the winding pack of 4 T. The
L-coil and C-coil have each been independently tested. The F-coil cryogenic test is under preparation. This
paper describes the results of the final assembly of the LDX experiment and the first integrated test of the Fcoil and the C-Coil in the normal LDX operating configuration.
Corresponding Author:
ALEX ZHUKOVSKY
zhukovksy@psfc.mit.edu
Massachusetts Institute of Technology, Plasma Science and Fusion Center, 77 Massachusetts Avenue, NW22,
Cambridge, MA 02139 U.S.A.
239
- E - Magnets and Power Supplies.
P2T-E-462
MODELING AC LOSSES IN THE ITER NBTI FULL SIZE JOINT
SAMPLES USING THE THELMA CODE
ZANINO ROBERTO, M.BAGNASCO 1, F.BELLINA 2, P.GISLON 3, P.L.RIBANI 4, L.SAVOLDI RICHARD 1
1 Dipartimento di Energetica, Politecnico, Torino, Italy 2 Dipartimento d’Ingegneria Elettrica, gestionale e Meccanica,
Università di Udine, Italy 3 ENEA, Frascati, Italy 4 Dipartimento di Ingegneria Elettrica, Università di Bologna, Italy
THELMA code is a tool for the numerical simulation of the behaviour of cable-in-conduit multistrand
superconductors (CICCs), like those to be used for the magnets of the International Thermonuclear
Experimental Reactor (ITER). Compared to other similar codes, the peculiarity of THELMA is the possibility
to analyse long CICC lengths, electrically connected each other by means of resistive joints, as foreseen for the
ITER magnets. The model solution is obtained solving simultaneously the electromagnetic and the thermalhydraulic coupled problems. THELMA is presently under debugging phase, and its first results have been
obtained and compared with experimental results. The paper presents the results of the analysis of two NbTi
Full-Size Joint Samples, tested at the Sultan facility of PSI in Villigen (CH). The first is the “PF-FSJS”, tested
in 2002, and the second is the “PFCI-FSJS”, tested in 2004, in the framework of the ITER R&D activities.
Both the samples are made of two “legs” (two straight parallel vertical pieces of CICC), each approximately
3.5 m long, electrically connected in series at their bottom with a resistive joint. The samples are cooled with
super-critical helium in forced convection at typically 5 K and 1 MPa. In the PF-FSJS the flow, in the whole
CICC cross section, is directed downwards, whereas in the PFCI-FSJS, the flow is restricted to the annular
region and directed upwards. For each sample, the two legs are of a different type: in the PF-FSJS the main
difference consists of the type of strand, while in the PFCI-FSJS the major difference is the presence or the
absence of wrappings around the CICC subcables. Both the samples are fully instrumented with magnetic
field, voltage, and temperature sensors. Therefore an exhaustive comparison between the experimental and the
computed quantities is possible. The tests are mainly aimed at the CICC and the joint characterization. We
concentrate here on the measurement of the AC losses, induced in the cable by the magnetic field generated by
a pulsed coil. In the THELMA model, one leg is considered, and the CICC is represented as 6 cable elements
corresponding to the CICC petals. The magnetic coupling between the two legs and with the AC field coils is
taken into account. In the paper, the distribution temperature and losses are reported as a function of time and
space, and the comparison with the experimental results is presented.
Corresponding Author:
ZANINO ROBERTO
roberto.zanino@polito.it
Dip. di Energetica, Politecnico, Troino, I-10129 Italy
240
- E - Magnets and Power Supplies.
P2T-E-471
POWER DISSIPATION AND ENERGY TRANSFER DURING TESTING
OF THE ITER TOROIDAL FIELD MODEL COIL
MARCHESE VITO, W.H. FIETZ, R. HELLER, M. SÜSSER, F. WÜCHNER, G. ZAHN
The test of the ITER Toroidal Field Model Coil (TFMC) in the background field of the EURATOM-LCT coil
took place in autumn 2002 at the TOSKA facility of the Forschungszentrum Karlsruhe in the framework of the
ITER R&D programme. The maximum currents in the two coils, in combined operation, were 16 kA in the
LCT coil and 80 kA in the TFMC respectively. Eddy currents are generated in the passive structures (e.g.,
stainless steel radial plates and coil cases) which are magnetically coupled to the coil windings, leading to
eddy current losses during current ramp up and down, as well as during a fast discharge. Due to the ripple of
the power supplies, based on twelve pulse ac-dc thyristor converters, losses were also generated during flat
top. Two He refrigerators of 2 kW and 0.5 kW respectively were used for the primary loop of the test
configuration. The heat load of both coils, including the eddy current losses in the passive structures and the
Joule losses due to the joint resistances, was removed by a secondary loop of forced flow supercritical He.
Both the TFMC and the LCT coil windings were cooled in series with their respective coil cases and Inter Coil
Structure (ICS) in order to reduce the overall mass flow rate. About 2% of the stored energy was transferred to
the cryogenic system after all the safety discharges and quenches of both coils together. Most of the energy
(about 98%) was extracted and transferred to the dump resistors of both coils, located outside the vacuum
vessel. The evaluation of the eddy currents and the power losses in the two coil cases and in the TFMC radial
plates, the latter one cooled indirectly through the TFMC windings, for different operating conditions, has
been performed with a SIMULINK computer code based on the full inductance and resistance matrices. In
addition to the two coil windings, a linear model of the power supplies and their current controllers and the
dump resistors was applied. The program computes also the heat load of the secondary loops based on a
simplified thermo hydraulic model, incorporating energy conservation and the heat transfer equations (0D),
whose parameters have been validated experimentally with calorimetric measurements. The program has been
used to evaluate the energy losses transferred to the cryogenic plant and to the external power circuit for the
simultaneous ramping down of the currents in both coils and for the safety discharge.
Corresponding Author:
MARCHESE VITO
vito.marchese@itp.fzk.de
Institute für Technische Physik, Forschungszentrum Karlsruhe GmbH, P.O. Box 3640, D - 76021 Karlsruhe,
Germany
241
- E - Magnets and Power Supplies.
P2T-E-490
DC AND TRANSIENT CURRENT DISTRIBUTION ANALYSIS FROM
SELF-FIELD MEASUREMENTS ON ITER PFIS CONDUCTOR
FORMISANO ALESSANDRO, T. BONICELLI (2), YU. ILYIN (3), A. NIJHUIS (3), C. MARINUCCI (4), R.
MARTONE (1), L. MUZZI (5), M. POLAK (6), C. SBORCHIA (2), B. STEPANOV (4), S. TURTÙ (5)
(2) EFDA Close Supp. Unit, Garching, Germany. (3) Univ. of Twente, Enschede, the Netherlands. (4) EPFL-CRPP, Villigen
PSI, Switzerland. (5) ENEA Frascati, Frascati, Italy. (6) Inst. Electr. Eng., Slovak Academy of Sciences, Bratislava, Slovakia.
Nowadays a number of studies are being performed to investigate actual Cable-in-Conduit Conductors (CICC)
behaviour under condition of practical interest for the ITER (International Thermonuclear Experimental
Reactor) magnets. Many aspects of such conductors are under examination, among which the impact of
working conditions on the current distribution among the conductor sub-structures. Unfortunately, direct
measurements of the current profile inside the conductor are not possible. An indirect approach for the current
profile estimate could be based on the measurement of the magnetic self-field around the CICC. The current
distribution should then be reconstructed using inverse problems methodology. In the present work, two
approaches to such problem are presented and compared. The first model adopted in the inverse problem
formulation is based on magnetostatic equations, and is able to take into account the internal structure of the
cable (3D), as well as the effect of external stray fields and other disturbance sources. In contrary, the second
model uses somewhat simplified geometry (2D) of the conductor, thus saving time for modelling and
computation. A comparison between two models should reveal an impact of the accuracy in the geometrical
representation of a cable on the result of the current reconstruction. The performance of the measurement
system is crucial, due to the extremely ill conditioning of the inverse problem; consequently, Hall Probes
assemblies (called “heads”) must be optimally designed and placed around the cable, in order to reach a
satisfactory trade-off between sensitivity and robustness in the measurement process. Some considerations
about the Hall sensors quality and the impact of geometrical uncertainties on the reconstruction process are
also discussed in the paper. The reconstruction procedures have been applied on the PFIS (Poloidal Field
InsertConductor Sample) tested in the SULTAN test facility under various working conditions. Three Hall
Probe heads have been placed on the conductor sections being in a high field region and in a low field one.
Details about the experiment and the results of the current reconstruction by two analytical approaches are
presented and discussed.
Corresponding Author:
FORMISANO ALESSANDRO
a.formisano@unina.it
Dip. di Ingegneria dell’Informazione, Seconda Universita' di Napoli, Via Roma 29, I-81031, Aversa (CE), Italy
242
- E - Magnets and Power Supplies.
P2T-E-491
THE MAGNET SYSTEM OF THE KTM TOKAMAK
BONDARCHUK EDUARD, E.N. BONDARCHUK (1) A.A. MALKOV (1) V.A. KOROTKOV (1) S.A. KRASNOV (1)
Y.M. KRIVCHENKOV (1) V.A. KRYLOV (1) A.B. MINEEV (1) A.K. CHERDAKOV (1) E.A. AZIZOV (2) V.N.
DOKOUKA (2)
(1) D.V. Efremov Institute, STC ‘Sintez’, Metallostroy, Doroga na Metallostroy 1, 196641 St. Petersburg, Russia (2) Troitsk
Institute for Innovation’s and Thermonuclear Researches, 142190 Troitsk, Russia
Magnet system of the KTM tokamak has been designed to provide certain shape and evolution of single null
plasma with aspect ratio A = 2. It makes possible, from one hand side, to create a compact and, relatively,
cheap machine that enables to solve requested tasks, and, from another hand side, to get required physical
parameters of plasma-magnet configurations that are between values typical for classical (A ? 2.5) and, so
called, spherical (A ? 1.6) tokamaks. Magnet system is a resistive pulsed system and consists of 20 toroidal
field coils (TF), 6 poloidal field coils (PF) and central solenoid (CS). The PF coils are located inside the TF
system. To make possible the installation of the one-piece welded vacuum vessel the TF coils are designed to
have joints. The TF coils produce a field of 1 Tesla at the plasma center of 0.86 m. Maximum plasma current
is 0.75 MA. Pulse duration is around 4 seconds. The cooling of the windings is ensured by pumping water
through the cooling channels. Magnet system has been designed to operate for the various of loading
conditions including temperature distributions in windings due to Ohmic heating, weight loads,
electromagnetic forces acting at the normal operation as well as for the scenario of central or vertical plasma
disruption. Results of structural analysis of the magnet system show that stresses in the conductors and support
structures are within the allowable limits. The CS coil is 4 layer wound coil and has 8 cooling passes. To
satisfy the requirements on the mechanical strength the silver bronze is selected as the material for both the CS
and central part of the TF conductors. Magnet system has been designed to withstand 20000 full size pulses.
Corresponding Author:
BONDARCHUK EDUARD
bondar@sintez.niiefa.spb.su
Efremov Institute, STC 'Sintez', Metallostroy, Doroga na Metallostroy 1, P.O. Box 42, 196641 St. Petrsburg,
Russia
243
- E - Magnets and Power Supplies.
P2T-E-511
OPTIMISATION OF THE CURRENT DISTRIBUTION IN THE IGNITOR
POLOIDAL FIELD COILS AND EVALUATION OF THE COILS
TEMPERATURES AND RESISTANCE DURING THE REFERENCE
OPERATING SCENARIO
RITA CAMILLO, COCILOVO WALTER (1), CUCCHIARO ANTONIO (1), GALASSO GIUSEPPE (2), PIZZUTO
ALDO (1), RAMOGIDA GIUSEPPE (1), ROCCELLA MASSIMO (1), PROF. COPPI BRUNO (3)
(1) Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy (2) Ansaldo Ricerche, Corso Perrone
25, 16152 Genova (GE), Italy (3) MIT, 02139 Cambridge (Ma), USA
The Ignitor Poloidal Field Coil (PFC) system consists of 13 pairs of coils up down symmetric, closely
distributed around the plasma column. The main relevant components are a Central Solenoid (CS) that
includes 7 coil pairs located between the Toroidal Field Coils (TFC) and the Central Post (CP), and a set (6
pairs) of external coils. The high number of independent coil pairs guarantees robust elongated plasma
equilibria while optimising the flux coupling with the plasma. The balance between the PFC flux capability
and the plasma flux requirement is one of the most crucial issues in designing compact high field toroidal
machines. The high current values in the coils needed to assure the plasma flux requirements could produce
high forces and temperature increases on the coils resulting in significant stresses in the mechanical structure.
The design of the IGNITOR tokamak requires a careful analysis because of the structural performance of the
machine relies on an optimised combination of bucking, between TF coils and the central solenoid, and
wedging, among the toroidal magnet legs. In this context the evaluation of the poloidal fields temperature and
resistances is of relevant importance for the IGNITOR design, because the low initial temperature of the coils
(30 K) and the high magnetic fields involved increase the role played by the magneto-resistance effect, that
produces a significant temperature gradient in the central solenoid coils. To calculate the poloidal coils
temperatures and resistances, during the whole IGNITOR reference scenario has been used a 2D axisymmetric
integral code developed in ENEA and its results have been compared with those one achieved with the
MAXWELL FEM code. Our code calculates both electromagnetic forces and resistivity for every turn of the
coils as function of the magnetic field and temperature through all the expected current scenario, for the
reference IGNITOR discharge at 11 MA. It is then calculated the increase of temperature due to the Joule
effect to get the maximum temperature value for each coil. Due to the short duration of the IGNITOR
discharge the heat transfer mechanism can be approximated as completely adiabatic, obtaining so a
conservative approach. The code results has evidenced critical temperature values on the inner central coils,
suggesting the opportunity of a current density redistribution. This goal has been obtained using a grading
technique for such coils.
Corresponding Author:
RITA CAMILLO
rita@frascati.enea.it
Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy
244
- E - Magnets and Power Supplies.
P2T-E-528
SAFETY ASSESSMENT OF THE ITER COILS SYSTEM
RABOIN SERGE, J.-L. DUCHATEAU (1) AND EISS TEAM
(1) Association Euratom-CEA, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France
In the ITER experimental reactor, the thermonuclear plasma is magnetically confined by means of a complex
system of superconducting magnets. The use of the superconducting technology plays a pivotal role in the
viability of the fusion technology, such magnets allowing substantial gains in electrical power consumed in the
operational phase of thermonuclear reactors. Superconducting coils however lead to severe operational
constraints, imposing mainly to maintain a homogenous and very low temperature inside the whole volume of
each of the 4 ITER coil sets. Taking into account on the one hand, that the coils are the seat of strong
electromagnetic energy transfers which may endanger their structure, and on the other hand, that they are
necessarily very close to the plasma chamber, their implementation in the nuclear environment of ITER has
imposed a drastic level of reliability and safety. The design and the safety assessment of the ITER magnets
result of a systematic and thorough approach. Potential initiating faults of every main component (conductor,
coil, structure, electrical and cryogenic services, instrumentation and control) have been identified and a
phenomenological analysis has allowed to define a set of conservative sequences. The analysis of these
sequences shows a limited environmental potential impact, far below the safety guidelines.
Corresponding Author:
RABOIN SERGE
Serge.RABOIN@cea.fr
Direction de l’Énergie Nucléaire, CEA Cadarache, 13108 Saint-Paul-lez-Durance, France
245
- E - Magnets and Power Supplies.
P2T-E-534
WINDING MACHINES FOR THE MANUFACTURING OF
SUPERCONDUCTIVE COILS OF THE MAIN EUROPEAN FUSION
RESEARCH MACHINES
CAZZANIGA RODOLFO, R. CAZZANIGA (1) R. PENCO (2) C.E. D’URSO (2)
(1) TPA s.r.l. via Ettore Paladini 16, 23891 Barzanò (LC), Italy (2) ANSALDO Superconduttori, Corso Perrone 73r, 16152
Genova, Italy
The successful completion of large magnet projects passes through the development and application of nonconventional manufacturing processes. The conductor winding process is a difficult and delicate step for the
manufacturing of superconducting coils. It is one of the most challenging and advanced and has to be
especially tailored for each project. The first task required for the winding process is the carrying out of
unusual, large, mostly three-dimensional coils. The second is to maintain as far as possible the section of the
cable unaltered in order not to damage the strands and to maintain the design features of the cable. The third is
to assure the suitable repetitiveness and speed for an industrial process. The manufacturing solution is a system
of different machines linked and tuned together and specially designed for each coil. Each machine must be
previously tested. A tailored software assures the overall process control. TPA provided ANSALDO
Superconduttori with the winding systems for all the major European projects: TFMC (of NET), CMS (of
CERN), WENDELSTEIN (of Max Planck IPP). The considerable experience gained in this field by both
companies, TPA and ANSALDO Superconduttori, has been just acknowledged by the CERN with “The CMS
gold Award of the Year 2004”. The paper reports the progress of the winding technology. It describes the main
features of the winding machines, the main problems, the layouts of the systems used in the above-mentioned
projects and the new ideas for the forthcoming ones
Corresponding Author:
CAZZANIGA RODOLFO
info@tpabrianza.it
TPA BRIANZA Via Paladini, 16 23891 Barzanò (Lecco) Italy
246
- E - Magnets and Power Supplies.
P2T-E-539
A SUCCESS STORY: LHC CABLE PRODUCTION AT ALSTOM MSA
GRUNBLATT GERARD, P. MOCAER (1) C. KOHLER (1)
(1) ALSTOM 3 Av des trois chenes 90000 Belfort France
ITER ,when constructed ,will be the equipment using the largest amount of superconductor strand ever built
(Nb3Sn and NbTi). ALSTOM Magnets and Superconductors SA, “MSA” received in 1998 the largest orders
to date for delivery of superconductor strands and cables (3100 km of cables and various strands) for the Large
Hadron Collider being built at CERN. These orders to MSA correspond to more than 600 metric tons of
superconducting strand ,amount to be compared to around 600 metric tons of Nb3Sn strands and 250 metric
tons of NbTi strands necessary for ITER. Starting from small and short R&D programs in the early nineties,
MSA has reached its industrial targets and has as of March 2004 delivered more than 60% of the whole orders
with products meeting high quality standards. Production is going on at contractual delivery rate and with
satisfactory financial results to finish deliveries around mid 2005 We will explain how we succeed to
transform a “cottage industry” (25 people in 1997) to a very high “world class” production activity (170 people
in 2004). Main industrial problems now solved such as investments and industrial set up and ramp up to reach
plateau production as well as more technical problems closely linked to industrial ones such as multifilament
wire breaks , strand magnetization, coating process (0.15µm controlled ) and others will be addressed and the
various methods used to solve such problems will be reviewed.
Corresponding Author:
GRUNBLATT GERARD
gerard.grunblatt@powerconv.alstom.com
ALSTOM 3 Av des trois chenes 90000 Belfort France
247
- F - PLASMA FACING COMPONENTS.
P4C-F-8
TILES CHAMFERING AND POWER HANDLING OF THE MK II HD
DIVERTOR
SALAVY JEAN-FRANÇOIS, P. CHAPPUIS (1) P.J LOMAS (2) V. RICCARDO (2)
(1) CEA Cadarache, Direction des Science de la Matière, F-13108 St Paul Lez Durance, France (2) UKAEA, Culham
Science Centre, Abingdon, OXON, OX14 3DB, United Kingdom
The JET HD (High Delta) Divertor is an upgrade of the actual JET divertor consisting of two modified toroidal
segments which are: a new Load Bearing Septum Replacement Plate (LB-SRP) tile located at the septum
position, and a High Field Gap Closure (HFGC) tile protecting inboard diagnostic cabling. The aim of the
upgrade is to allow high power operation and a wider range of plasma lower triangularities. This paper
describes the optimisation of the tile chamfering for LB-SRP and HFGC (including edge shadowing) and of
the power handling evaluation for a set of planned plasma configurations. The LB-SRP and HFGC tiles have a
slope in the toroidal direction to hide any edge of the next tile from the impinging plasma. They are machined
from blanks of Carbon Fibre Composites materials and are attached to the carrier through the JET usual system
of dumbbell, tie rod and disc springs. The precise design of the tile faces is based on 12 plasma configurations
given by the JET team, and on two sets of mechanical tile tolerances, issued by the JET drawing office. The
PROTEUS code (magnetic equilibrium by finite element) is used to calculate the various field lines angles,
which are inputs for the chamfering angle calculation process. The design of the LB-SRP tile has been
optimised to increase the power handling (roughly from 1 to 80MJ) for the different plasma configurations and
for the various tile tolerances (chamfering angle varies linearly along the tile profile). After calculating the
chamfering angle values of each face, a checking exercise has been realised on the 3D CATIA models of the
tiles by putting them at their extreme tolerance positions and validation if the shadowing is insured for a angle
calculated to take into account the worst possibilities. With the final chamfering angle value for each face, the
power handling of the tiles has been estimated with final elements calculations. Power handling are given
either with the critical time to reach 1800 C at the tile surface for a total injected power of 40 MW, or with the
maximum total injected power allowable for a 10 seconds run without reaching 1800 C. Chamfering angles
for the new LB-SRP and HFGC tiles of the MK II HD divertor have been calculated, optimised and checked in
order to insure a good shadowing of the edges for each of the 12 reference plasma configurations. The
consequent power handling has been estimated and gives promising results in regards to the JET EP project
objectives.
Corresponding Author:
SALAVY JEAN-FRANÇOIS
salavy@cea.fr
CEA Saclay, DEN/DM2S/SEMT/BCCR, F-91191 Gif sur Yvette, France
248
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P4C-F-12
THERMAL AND MECHANICAL ANALYSIS OF THE EAST PLASMA
FACING COMPONENTS
SONG,YUNTAO, D.M YAO, S.T WU, P.D WENG
Institute of Plasma Physics, Chinese Academy of Sciences P.O.Box 1126, Anhui, Hefei, P.R.China, 230031
The EAST (Experimental Advanced Superconducting Tokamak) is an advanced steady-state plasma physics
experimental device to be built in P.R.China. It has a long pulse (~1000s) capability and will be able to
accommodate divertor heat loads that make it an attractive test for the development of advanced tokamak
operating modes. Now, the engineering designs for the EAST plasma-facing components (PFC) are in
progress. The EAST PFCs consist of a plasma-facing surface affixed to a cooled support plate. All the PFCs
are made of copper alloys (CuCrZr) on which Carbon-carbon (C-C) composites tiles are mechanically
attached. A thin piece 0.3mm of flexible graphite sheet is used between the tile and the heat sink to improve
the contact conductance. This paper will introduce the thermal and mechanical analysis, which included the
structure intensity and optimization hydraulic parameters by the finite element method for the different type of
cooling and mounting structure between the heat sink and C-C tiles using for EAST plasma facing
components. Since EAST is a long pulse machine, all of the analysis was done for steady-state conditions. To
reduce the radiation enhanced sublimation, the maximum temperature are controlled to be less than 1000Ž.
In order to reduce the impurity sources to plasma, the minimum wall temperature has to be more than
100Ž.The maximum heat flux 7MW/m2 on these components is chosen based on the simulation of the
plasma operation. In this study three types of PFC mounting structures are considered: 1) the C-C tiles bolted
to the water-cooled copper plates through the thin piece of flexible graphite sheet. The copper alloys are
actively cooled by water flowing the cooling channel drilled by a special punch. 2) C-C tiles bolted to the
copper plates through the thin piece of flexible graphite sheet and welding a water-cooling tube on the other
side of copper plates. 3) C-C tiles bolted to the copper plates, which brazing on a water-cooled and supporting
stainless steel structure. Based on the optimization analysis results the type 1 is chosen as the mounting
structure between the tiles and heat sinks for the PFCs of EAST device, the maximum water-cooling velocity
chose as 7m/s. Under these conditions the maximum thermal stresses in copper plate is 220MPa,which is less
than the allowable stress based on the design criteria ASME code. The maximum temperature of C-C tiles is
829Ž, which also have been proved by a prototype test.
Corresponding Author:
SONG,YUNTAO
songyt@ipp.ac.cn
P.O.Box 1126, Anhui, Hefei, P.R.China, 230031,Institute of Plasma Physics, Chinese Academy of Sciences
249
- F - Plasma Facing Components.
P4C-F-13
THE DYNAMIC ERGODIC DIVERTOR IN TEXTOR – A NOVEL TOOL
FOR STUDYING MAGNETIC PERTURBATION FIELD EFFECTS
O. NEUBAUER, B. GIESEN, P.W. HÜTTEMANN, H.T. LAMBERTZ AND THE TEXTOR TEAM
Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, Association EURATOM/FZJ, Trilateral Euregio Cluster, D52425 Jülich, Germany
Recently TEXTOR has been upgraded by installation of the Dynamic Ergodic Divertor (DED). The purpose of
the DED is to influence transport parameters in plasma edge and core and to study the resulting effects on heat
exhaust, edge cooling, impurity screening, plasma confinement and stability. Alternatively, the DED creates
static or rotating multipolar helical magnetic perturbation fields of different mode patterns. A set of 16 helical
coils has been installed on the inboard high-field side of the vacuum vessel covering about one third of the
plasma surface. Thus, in contrast to similar experiments, the DED produces a clear mode spectrum with mode
numbers of m/n = 12/4, 6/2, 3/1 or a superposition of those. For the first time rotating fields of up to 10 kHz
can be generated. The peak coil current is 15 kA for pulse duration of up to 10 s. A novel coil design has been
developed which fulfils the various mechanical, electrical, high frequency, thermal, and vacuum requirements.
The coils consist of twisted copper wires in a corrugated stainless steel tube. A combined Helium / water
cooling system removes the energy of adiabatic heating of the coils during a pulse. Coaxial vacuum
feedthroughs have been designed which allow for compensated current feeding as well as supply of the
cooling media. During a major shut down the DED components have been integrated. For this purpose, after
removing diagnostics, TEXTOR has been split into two parts; the liner has been removed, modified and
reintegrated, followed by the mounting of DED components. In parallel the power supply system has been
fabricated, installed and tested on a full size dummy load. Particularly the AC operation has been realised by
modular IGBT series resonant inverters. A sophisticated control system guarantees sufficient stability of
amplitude, frequency and phase of the coil currents. In addition to the various technical aspects of the DED
design, implementation and commissioning, highlights of recent experiments will be presented. In particular
the impact of the perturbation field on MHD stability and plasma rotation will be addressed.
Corresponding Author:
O. NEUBAUER
o.neubauer@fz-juelich.de
Forschungszentrum Jülich GmbH, IPP, 52425 Jülich, Germany
250
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P4C-F-24
EAST(HT-7U) IN-VESSEL COMPONENTS DESIGN
YAO DAMAO, J.G.LI; Y.T.SONG; S.J.DU; J.L.CHEN; P.D.WENG
P.O.Box1126 Hefei Anhui 230031, P.R.China
In vessel components are very important parts of EAST (HT-7U) superconducting tokamak. The primary
purpose of these components is to protect the vacuum vessel, RF system and diagnostic components from the
plasma particles and heat loads. Other function of in vessel components is additional to particles and heat loads
management. The divertor is designed to provide particles exhaust into the divertor cryopump, provide
recycling control, and impurity control. The passive plates stabilize the plasma vertical stability. Heat loads
and electric-magnetic forces are quite complicate for in-vessel components. Considering the possibility of
plasma operation condition for EAST different operation phases, the initial operation of EAST is focus on
physics scenarios, plasma elongation experiment explore and plasma control experiment explore etc. The long
pulses and steady state operation is planed in the later time of first phase. During the first stage plasma heating
power will be reached 10MW, and maximum heat flux will not permit more than 1MW/m2 for steady state
operation. Brazed tile are not employed for divertor plate. All boron doped graphite tiles with SiC coating are
bolted to active water-cooling CuCrZr heat sink. The bolted structure is also used for inner toroidal limiter and
passive stabilizer to handling heating power up to 10MW when the plasma is operated in circle cross-section.
EAST divertor layout is designed as up-down symmetry to accommodate both double null and single null
plasma configuration. It is provide a large experimental flexibility and capable of running in a stand scenario,
with power conducted along the field lines to the target plates, or in a radiative divertor mode. The geometry of
the divertor is based on simulations obtained using the EFIT code and references the experiences of physic
design, engineering design and experiment of other tokamak. The vertical target is inclined so as to intercept
the magnetic field lines of the separatrix at an acute angle. A “V” shape was formed and expect particles
remain in this region to aids heat load uniform distribute on divertor plate so as to reduce the peak heat flux on
divertor targets. Passive stabilizer is a single turn saddle coil with active water cooling and a single vertical
current bridge. Consider the flexibility of plasma control several copper saddle coils will be used for plasma
equilibrium control, error field corraction and RWM control.
Corresponding Author:
YAO DAMAO
yaodm@ipp.ac.cn
P.O.Box1126 Hefei Anhui 230031, P.R.China
251
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P4C-F-27
HOT RADIAL PRESSING: AN ALTERNATIVE TECHNIQUE FOR THE
MANUFACTURING OF PLASMA-FACING COMPONENTS
VISCA ELISEO, S. LIBERA (1), G. MAZZONE(1), A. PIZZUTO(1) C. TESTANI (2)
(1)Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, IT-00044 Frascati (2)CSM S.p.A., IT-00128 Castel Romano,
Roma
The Hot Radial Pressing (HRP) manufacturing technique is based on the radial diffusion bonding principle
performed between the cooling tube and the armour tile. The bonding is achieved by pressurizing the cooling
tube while the joining interface is kept at the vacuum and temperature conditions. This technique has been
used for the manufacturing of relevant mock-ups of the ITER divertor vertical target. Tungsten monoblock
mock-ups were successfully tested to high heat flux thermal fatigue (18 MW/m2 for 1000 cycles). After these
good results the activity is now focused on the developing of a canister suitable for the CFC monoblock mockups. The stainless steel canister reutilization is also one of the main objectives. A FE calculation was
performed to investigate the stress involved in the CFC tiles during the process and to avoid the CFC fracture.
The influence of the process parameters was also investigated in order to keep the process bonding
temperature as low as possible to preserve the copper alloy thermo-mechanical properties. The design
improvement of the canister and the equipment will be reported in the paper as well as the results obtained by
FE calculation and by the HRP manufacturing of the monoblock mock-ups.
Corresponding Author:
VISCA ELISEO
visca@frascati.enea.it
Associazione EURATOM-ENEA sulla Fusione - Via E. Fermi, 45 - 00044 Frascati RM
252
- F - Plasma Facing Components.
P4C-F-28
HETS PERFORMANCES IN HE COOLED POWER PLANT DIVERTOR
ALDO PIZZUTO, PANOS KARDITSAS (+), CLAUDIO NARDI (*), STAMOS PAPASTERGIOU (X)
(*) ENEA – CR Frascati – Via E. Fermi 45 – I-00044 Frascati (Roma) Italy (+) UKAEA – Culham Science Centre,
Abingdon, Oxfordshire – OX14 3DB, UK (x) EURATOM at ENEA Frascati
In the frame of the activities aimed to evacuate the performances of a He cooled divertor in the future fusion
power plant, the High Efficiency Thermal Shield (HETS) concept has been proposed. This concept relays on
an abrupt change of momentum of the fluid in order to increase the turbulence in the gas, and therefore the
heat transfer. This concept, initially developed for water, has been extended in the past years to He, and studies
have been performed in order to evaluate his suitability in such environment. The requirements for the power
plant divertor are to sustain a thermal flux of at least 10 MW/m2, without exceeding limits in stress and
temperature. A further limitation is given by the required pumping power, because the attractiveness of the
power plant is strictly related to the power output of the plant itself. As the PPS divertor must operate with an
energy production plant, the He must have such a characteristics to be used directly in the energy production
cycle (gas turbine), therefore the reference values are of 10 MPa He pressure and an inlet temperature of 600 C
and outlet temperature 800 C (temperature rise of 200 C, this value could be re-evaluated in order to optimize
the characteristics. Analytical studies developed in ENEA and UKAEA showed that the HETS concept can
sustain a thermal flux of 10 MW/m2, keeping low pressure drops (and therefore pumping power). A thermal
flux as high as 15 MW/m2 can be easily sustained, increasing the pumping power up to the proposed limit
(10% of the thermal power to the divertor). Because of the relevance of pressure drop in the structure for the
performances of the divertor, experimental validations are required for the HETS, as in literature reliable
assessments for this characteristic can not be found. The preliminary experiments have been performed using
air at room temperature and high pressure, showing values of pressure drop in line with the parameters used in
the calculations. Studies aimed to optimize the shape of the channel, in order to further reduce the pressure
drop are in progress. At the present stage the HETS appear to be a promising solution for the He cooled power
plant divertor.
Corresponding Author:
ALDO PIZZUTO
pizzuto@frascati.enea.it
ENEA - Via E. Fermi 47 - 00044 Frascati (Roma)
253
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P4C-F-32
THE INFLUENCE OF IRRADIATION REGIMES ON RETENTION
HYDROGEN ISOTOPES IN STRUCTURAL MATERIALS
ZALUZHNYI ALEXANDER GEORGIEVICH, KOPYTIN VLADIMIR PLATONOVICH SUVOROV ALEXANDER
LEONIDOVICH
B. Cheryomushkinskaya 25, Moscow 117259, Russia
Researching of the hydrogen isotopes retention in candidate materials, which connects to possible loosing of
valuable fuel (tritium) and service safety of fusion reactor, becomes very important because of elaboration of
the first-wall materials for fusion reactor. In the present work was investigated the influence of irradiation
regimes on retention hydrogen isotopes in samples of austenitic steel during heating. The samples of studied
materials were irradiated both in the reactor and by hydrogen isotopes ions of different energies and fluencies
bombardment in an accelerator. Kinetic of hydrogen release from the samples worked with deuterium plasma
was investigated. The following results were obtained. Heating the irradiated samples of steel (irradiated in the
reactor or by hydrogen isotopes ions bombardment), which have been kept in normal temperature during quite
a long period after the irradiation, a shift of the diffusion peak of hydrogen release to higher temperatures,
comparing to no irradiated samples, was observed. It means that atoms of hydrogen in the irradiated sample
were caught by radiation defects, which are very effective as traps for hydrogen atoms till quite high
temperatures (700 K). The worked out analysis of the received results supposes that vacancy complexes. On
thermodesorption curves of hydrogen release from irradiated samples of austenitic steels a high temperature
peak (900-1000 K) was observed because of dissociation of hydrogen containing compounds in micro pores.
During investigations of hydrogen release from irradiated samples of austenitic steel, after it had been
saturated with hydrogen plasma, abnormally big blisters were registered with cover thickness of about 1mkm.
Three peaks were observed on the thermodesorption curves of hydrogen release from irradiated samples,
contained blisters. The low temperature spike (~500 K) was showed to correspond to hydrogen release because
of its resolution from blisters, where it was in molecular form. The high temperature peak (?900 ?) corresponds
to hydrogen release from dissociating blisters, which contain hydrocarbons. The mechanism of abnormal
blisters generation is offered. Inasmuch methane is not soluble in metals in temperatures lower then
temperature of its dissociation, it behaves as a noble gas, when heated to temperatures lower then temperature
of its dissociation.
Corresponding Author:
ZALUZHNYI ALEXANDER GEORGIEVICH
zaluzhnyi@itep.ru
Kashirskoe Shosse 31, Moscow 115409, Russia
254
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P4C-F-33
MANUFACTURING TECHNOLOGY DEVELOPMENT FOR THE
VACUUM VESSEL AND PLASMAFACING COMPONENTS
LAITINEN ARTTU, JARI LIIMATAINEN * MIKA KORHONEN * PENTTI HALLILA * SEPPO TÄHTINEN **
* Same address as for the corresponding author ** VTT Industrial Systems, P.O.Box 1704, FIN 02044 VTT Finland
Vacuum vessel and plasma facing components of the Iter construction including shield modules and primary
first wall panels have great impact on the production costs and reliability of the installation. From the
manufacturing technology point of view, accuracy of shape, properties of the various austenitic stainless
steel/austenitic stainless steel interfaces or CuCrZr/ austenitic stainless steel interfaces as well as stress
corrosion and fatigue properties of the base materials are crucial for technical reliability of the construction.
The current approach in plasma facing components has been utilization of solid-HIP technology and solidpowder-HIP technology. Due to the large size of especially shield modules shape control of the internal
cavities and cooling channels is extremely demanding. This requires strict control of the raw materials,
especially powder size distribution, encapsulation technology and hot isostatic pressing pressing cycle forms.
For the vacuum vessel components there are several optional manufacturing routes including narrow gap and
beam welding (laser or electron beam) of the bended thick walled plates and forgings, or optionally use of
forgings and thich walled plates together with hot isostatically pressed components in order to reduce total
amount of welding and to be able to position welds in a way that minimizes distortion risks during assembly
welding. In this presentation, different manufacturing methods area compared and their relative attractiveness
is evaluated and discussed.
Corresponding Author:
LAITINEN ARTTU
arttu.laitinen@metso.com
Rieväkatu 2, P.O.Box 1100, FIN-33541 Tampere Finland
255
- F - Plasma Facing Components.
P4C-F-41
ENGINEERING AND THERMAL-HYDRAULIC DESIGN OF PFC
COOLING FOR SST-1 TOKAMAK
CHAUDHURI PARITOSH, P. SANTRA, N. RABI PRAKASH, S. KHIRWADKAR, G. RAMASH, S. DUBEY, A. ARUN
PRAKASH, D. CHENNA REDDY, AND Y. C. SAXENA
Institute for Plasma Research, Bhat, Gandhinagar, Gujarat, INDIA
Steady state Superconducting Tokamak (SST-1) is a medium size tokamak with superconducting magnetic
field coils. It is a large aspect ratio tokamak with a major radius of 1.1 m and minor radius of 0.20 m. SST-1 is
designed for plasma discharge duration of ~1000 seconds to obtain fully steady state plasma with total input
power upto 1.0 MW. PFC is one or the major sub-systems of SST-1 tokamak consisting of divertors, passive
stabilizers, baffles, and poloidal limiters are designed to be compatible for steady state operation. The main
consideration in the design of the PFC is the steady state heat removal of upto 1MW/m2. In addition to remove
high heat fluxes the PFC are also designed to be compatible for high temperature baking at 350 C. Extensive
studies, involving different flow parameters and various cooling layouts have been examined to select the final
cooling parameters and layout, compatible for cooling and baking. Design considerations included 2-D steady
state and transient thermal analysis of PFC during plasma operation. Thermal analysis is carried out with the
purpose of evaluating the thermo-mechanical behavior of the PFC. Both 1-D analytical and 2-D Finite Element
(FE) analysis were carried out to determine the temperature distribution and the thermally induced stresses in
PFC. The results of the calculation led to a good understanding of the temperature and thermal stress
distribution in various parts of the PFC. Since the tiles are mechanically attached to the back plate, the fitting
technique must provide the highest mechanical stress so that thermal transfer efficiency is maximized. Proper
brazing of cooling tube on the copper back plate is necessary for the efficient heat transfer from the tube to the
back plate. The contact at the brazed joint of the tube to the backplate is critical for the above application.
Using an infra-red-camera, spatial and temporal evaluation of the temperature profile has been studied under
various flow parameters to evaluate the quality of the brazed joint of the manufactured modules. The
optimized thermal-hydraulic design and the effect of stress and strain on different material used in PFC were
examined and discussed in this paper.
Corresponding Author:
CHAUDHURI PARITOSH
paritosh@ipr.res.in
Institute for Plasma Research, Bhat, Gandhinagar, Gujarat, INDIA
256
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P4C-F-49
THE USE OF COPPER ALLOY CUCRZR AS A STRUCTURAL
MATERIAL FOR ACTIVELY COOLED PLASMA FACING AND IN
VESSEL COMPONENTS
LIPA MANFRED (1), A. DUROCHER (1), R. TIVEY (2), TH. HUBER (3), B. SCHEDLER (3), J. WEIGERT (4)
(1) CEA/Cadarache, DRFC, F-13108 SAINT PAUL LEZ DURANCE, France (2) ITER Garching Joint Work Site, D-85748
Garching, Germany (3) Metallwerk Plansee GmbH, A-6600 REUTTE, Austria (4) PTR Präzisionstechnik GmbH, D-63461
Maintal-Doernigheim, Germany
Within the European fusion community there is a wide experience of the use of copper chromium zirconium
(CuCrZr). Precipitation hardened CuCrZr has been used as structural material for actively cooled plasma
facing components for 15 years in Tore Supra (TS), and more than 20 years in JET. In the future CuCrZr will
be used for the actively cooled divertor plates of the W7X stellarator and has been selected as the heat sink
material for the ITER divertor. In all cases TS, JET, W7X and ITER the components are cooled using
pressurised hot water and as they operate in ultra high vacuum have to remain leak tight in operation, during
which they are exposed to cyclic thermal loads, and in the case of TS and ITER large electro dynamic forces,
and specifically for ITER neutron irradiation. The CuCrZr pre-material for TS and JET component fabrication
has been procured from several suppliers and delivered in product shapes such as: drawn rods, bars, hollowprofiles and in forged plates or rolled sheets. In the case of TS the material has always been delivered in the
solution annealed, quenched and age hardened state, whereas JET prefers to use solution annealed material
only performing hardening operations towards the end of the fabrication process. This paper presents our
experience with this material in the procurement and fabrication of actively cooled components for TS (pump
limiter fingers, guard limiter heat sink hollow-profiles, ripple protection tubes, endoscope heads), JET
(accelerations grids and beam scrapers for the beam-lines) and divertor prototypes tested for ITER. By
highlighting failed manufacturing processes it is hoped that future users can avoid repetition of costly and
time-consuming failures that might occur both during manufacture and during subsequent operation. To this
end the technical specification for procurement of CuCrZr is discussed including the supposed influence of
chemical impurities, alloying elements, material production process and heat treatments on material properties.
The general properties of different material grades, procured from different suppliers for various component
applications, are also given. The associated mechanical characterisation of component joining, focused on
fusion welding using electron beams both of CuCrZr to SS via a nickel adaptor and CuCrZr to CuCrZr, is
presented. The behaviour of actively cooled CuCrZr components during normal and especially during
accidental operation conditions is described.
Corresponding Author:
LIPA MANFRED (1)
manfred.lipa@cea.fr
CEA Cadarache, DRFC, 13108 Saint Paul lez Durance, France
257
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P4C-F-54
MANUFACTURING OF THE W7-X DIVERTOR AND WALL
PROTECTION
STREIBL BERNHARD, J. BOSCARY, P. GRIGULL 2), H. GREUNER, J. KIßLINGER, C. LI, B. MENDELEVITCH,
T. PIRSCH, N. RUST, S. SCHWEIZER, M. WEIßGERBER
Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, D-85748 Garching 2) Max-Planck-Institut für
Plasmaphysik, Euratom Association, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald
The W7-X stellarator is designed for steady state operation with an input power of 10 MW and transient
discharges up to 20 MW for 10 seconds. In the first step an 'open divertor' will uncouple the plasma core from
the wall. According to the structure of the plasma boundary 10 water-cooled divertor units, one lower and one
upper per field period, are arranged such that leading edges are avoided. Each divertor unit, composed of
horizontal and vertical modules adjacent to the pumping slot, is 5 m long and spans over 70% of the field
period. Baffle modules increase the neutral density in front of the target plates. To improve the particle control
a closed divertor chamber is formed by toroidal and poloidal end plates and two cryo-pump units behind the
horizontal target increase the pumping speed. Behind the vertical baffle a control coil is arranged for
compensating symmetry breaking error fields and sweeping the strike points. The remaining poloidal and
toroidal area of the plasma vessel is covered by a water-cooled wall protection. On the outboard, where the
lowest heat flux is expected, steel panels are applied. They protect also the access ports over the length of their
typical diameter. On the inboard side of the plasma vessel heat sinks armoured with clamped graphite tiles are
arranged. The same design is applied for the baffle modules and two of the 9 horizontal target modules with a
considerably reduced heat load of 1 MW/m2. Only the width ratio of heat sinks and tiles is adjusted
appropriately to the heat flow. The remaining target modules are composed of 6 to 12 elements, equipped with
CFC tiles of Sepcarb® NB31 to take up 10 MW/m2. In total 940 elements of this kind, including 50 spares,
are required and will be manufactured by the company Plansee. Profit will be taken from the CIEL experience
via the collaboration with CEA. The company MAN-DWE will manufacture the steel panels. Also an external
company produces the 10 control coils. By IPP itself, the modules with clamped tiles and the cryo-pumps will
be manufactured and qualified with pre-series tests. In addition IPP will perform all acceptance tests: the high
heat flux tests for the target elements with an ion beam facility, the vacuum and out gassing tests in a large
vacuum oven, the water flow and pressure tests and the final tests of the control coils at nominal electrical
current. Special technological aspects will be qualified in close collaboration between the manufacturers and
IPP.
Corresponding Author:
STREIBL BERNHARD
Bernhard.Streibl@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, D-85748 Garching
258
- F - Plasma Facing Components.
P4C-F-59
STUDIES ON GRAPHITE SURFACES DETRITIATION BY PULSED
REPETITION RATE NANOSECOND LASERS
SEMEROK ALEXANDRE, WEULERSSE J.-M., BRYGO F., LASCOUTOUNA CH., HUBERT CH. , LE
GUERN F., *TABARANT M.
CEA Saclay, DPC/SCP/LILM, 91191 Gif sur Yvette CEDEX, France *CEA Saclay, DPC/SCP/LRSI, 91191 Gif sur Yvette
CEDEX, France
The detritiation of plasma-facing components is regarded as one of the crucial problems in a future thermofusion reactor design and construction. A fast heating of an exposed surface by a focused laser beam allows to
obtain the temperature higher than 1000K on a thin (1 – 100 µm) near-surface layer, thus, resulting in
detritiation either by hydrogen desorption from the surface or by ablation of this near-surface layer. Thus,
visible and near-IR lasers with 100-500W average power that can be transmitted by optical fibers may provide
a completely automatic unattended system for reactor vacuum chamber surfaces in situ cleaning. Low and high
repetition rate nanosecond laser benches provided with a sufficiently complete control and measurement
equipment were developed and applied for studies on graphite and co-deposited layer heating and ablation.
Heating and ablation regimes of detritiation (dehydrogenization) with pulsed lasers were distinguished by
ablation threshold fluency that was determined experimentally for graphite samples with D/H isotopes from
TexTor and TORE SUPRA (CEA Cadarache). Ablation threshold fluencies were Fth = 0.4 ± 0.1 J cm-2 for a
co-deposited layer and Fth = 1 J cm-2 for graphite surfaces.For graphite samples from TORE SUPRA and
TexTor, the ablation efficiencies were determined to be different: (0.025 µm/J cm-2) for graphite and (0.2
µm/J cm-2) for co-deposited layer. The particular features of the graphite and co-deposited layer ablation
(different ablation thresholds and laser ablation efficiencies) are discussed with relation to the procedures that
could ensure self-controlled laser surface cleaning. As the result of our investigations, the conclusion was
made that detritiation rate of 1 m2 per hour can be obtained for 20 µm co-deposited layer with high repetition
rate Nd-YAG laser beam of 250 W mean power. In this case, the laser fluency should be 1 J/cm2 to provide
the maximum ablation efficiency of 0.2 µm/Jcm-2. At the same time, we concluded that detritiation of a thick
co-deposited layer by laser heating is much more efficient with a continuous wave Nd-YAG laser radiation.
Experimental and theoretical studies on laser heating and ablation with different types of co-deposited layers
and Nd-YAG lasers (cw and pulsed) in controllable environmental conditions (gas composition, pressure) are
in progress and will be presented in the paper.
Corresponding Author:
SEMEROK ALEXANDRE
asemerok@cea.fr
CEA Saclay, DPC/SCP/LILM, B.467, 91191 Gif sur Yvette CEDEX France
259
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P4C-F-66
STEADY STATE AND TRANSIENT THERMAL-HYDRAULIC
ANALYSES ON ITER DIVERTOR MODULE
DELL'ORCO GIOVANNI *, ANCONA ANTONELLA**, DI MAIO PIETRO ALESSANDRO**, MEROLA MARIO***,
VELLA GIUSEPPE**,
* ENEA, P.O. Box 1, 40032 Camugnano (Bo) Italy, ** Università di Palermo - DIN, V.le delle Scienze, 90128 Palermo Italy,
*** EFDA CSU Garching Boltzmannstr. 2, D-85748 Garching-Germany
The Divertor is one of the most challenging components of the next step ITER nuclear fusion reactor. It is
aimed at reducing the impurities in the plasma and at sustaining the heat and particle fluxes during normal and
transient operations as well as during disruption events. The ITER Divertor consists of 54 cassettes and three
plasma-facing components (PFCs), namely the inner vertical target, the outer vertical target and the dome
liner. The water maximum total flow rate should be 1000 kg/s, with 100-150 C inlet/outlet temperatures, 4.2
MPa inlet pressure and a maximum pressure drop of 1.4 MPa. The PFCs are cooled in series, with a maximum
water velocity in the channel of 11 m/s, whilst the water coolant is routed via the cassette body. Each PFC
consists in a number of plasma facing units, cooled in parallel and assembled onto a supporting structure. Due
to the extremely high heat loads expected onto the PFCs (up to 20 MW/m2 over 20 s), the hydraulic design of
the DIVERTOR is particularly demanding. It shall ensure that the foreseen flow rate actually reaches each
plasma-facing unit to ensure an adequate cooling and to prevent any risk of Critical Heat Flux (CHF).
Sufficient margin ( > 40 %) to avoid the reaching of a CHR limit on the PFCs could be obtained by using
hypervapotron design inside the flat channels and swirl flow turbulence tape promoters inside the cooling
tubes. Furthermore the overall pressure drop and flow rate shall be within the specified design limit to avoid an
unduly high pumping power. Another important issue is the definition of a proper procedure to drain the
coolant and dry the Divertor components prior to the maintenance operations as well as to refill them with
water after maintenance ensuring a complete elimination of gas bubbles. Due to the complex flow scheme of
the hydraulic circuit, a pure theoretical study does not appears sufficient to address all the above-mentioned
items and an experimental validation of the models is mandatory. In addition to that, the assembly of the PFCs
onto the cassette body as well as their integration by welding the coolant connections of the PFCs, also
represent a critical step to be investigated. The paper presents both the steady sate and transient theoretical
thermal hydraulic analyses, carried out by RELAP code, on the Divertor module for the: i) flow distribution,
pressure drop and Critical Heat Flux margin; ii) draining and drying of the Divertor components.
Corresponding Author:
DELL'ORCO GIOVANNI *
giovanni.dellorco@brasimone.enea.it
ENEA, P.O. Box 1, 40032 Camugnano (Bo) Italy
260
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P4C-F-69
APPLIED TECHNOLOGIES AND INSPECTIONS FOR THE W7-X PRESERIES TARGET ELEMENTS
BOSCARY JEAN, H. GREUNER K. SCHEIBER(1) B. STREIBL B. SCHEDLER (1) B. MENDELEVITCH J.
SCHLOSSER (2)
(1) Plansee Aktiengesellschaft, Technology Center, A-6600 Reutte, Austria (2) CEA Cadarache, Euratom Association, F13108 St Paul-lez-Durance, France
The WENDELSTEIN 7-X (W7-X) divertor is designed to remove 10 MW during steady state operation. 23
m2 of the target area are approximated by 890 water-cooled target elements of 14 various types. The watercooling characteristics are optimized to sustain 10 MW/m2 and to remove 100 kW maximum per element.
Results of finite element calculations show the maximum allowable transient up to steady state heat load for
standard and diagnostic elements. The manufacturing process uses well-known technologies but applied to the
particular W7-X geometry. All plasma facing target elements are covered with CFC Sepcarb® NB31 flat tiles
(standard tile). They are bonded to Cu by Active Metal Casting (AMC®). The heat sink is made of CuCrZr.
All these processes will be qualified during the pre-series phase. The characterization of the first delivered
CFC batch of 150 kg shows that the thermal and mechanical properties of blocks are compatible with the
AMC® process. More than 2/3 of the elements are shielded by a L-shaped front tile against 1 MW/m2. This
solution allows the 3D fitting along the divertor pumping space. 30 diagnostic elements are equipped with
lateral tiles specified for 3 MW/m2. These particular tiles require to try two technologies for the bonding of the
AMC®-NB31 tile to the heat sink, namely electron beam (EB) welding and hot isostatic pressing (HIP). On
one hand, HIP guarantees a better thermal contact of the tile to the heat sink and avoids the access difficulty of
EB-welding for lateral tiles. On the other hand, HIP has never been applied to such a large production. The
selection between the two technologies is one of the issues of the pre-series phase. The four pre-series
elements and relevant specimens manufactured with this aim will be described. Non-destructive examinations
of the bond between the tile and the heat sink are integrated throughout manufacturing and applied at a very
early stage of the fabrication. AMC®-NB31 standard tiles are examined by X-ray and lock-in thermography.
The bond of these tiles to the heat sink block is checked by ultrasonic and lock-in. Lock-in is finally applied
for the completed target element, equipped with cooling channels. The crucial issue of the pre-series phase is
the definition of the acceptance criteria applied to the series production.
Corresponding Author:
BOSCARY JEAN
jean.boscary@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, D-85748 Garching bei München, Germany
261
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P4C-F-76
OVERVIEW OF THE ENGINEERING DESIGN OF THE ITER
DIVERTOR
TIVEY RICHARD, V. CHUYANOV, E. D’AGATA, G FEDERICI, H. HEIDL, A. MAKHANKOV, M. MEROLA, J.
PALMER
ITER Joint Work Site, Boltzmannstr.2, 85748 Garching, Germany D.V. Efremov Research Institute, St. Petersburg, Russia;
EFDA Close Support Unit, Boltzmannstr.2, 85748 Garching, Germany
There have been significant developments in the design and R&D of the ITER divertor since it was last
reported in the ITER Final Design Report 2001 (FDR 2001). The main developments will be presented and
these are outlined below. The construction and testing of prototypical components has demonstrated the
capability of handling, with sufficient margin, the predicted steady state heat flux on carbon-fibre composite
(CFC) and tungsten-armoured surfaces. There has been feedback from European industry on the
manufacturing difficulties associated with building of these large-scale pieces. As a direct consequence, a
number of design options are under consideration aimed at simplifying manufacture and/or reducing costs.
These include the option to keep as separate pieces the CFC- and tungsten-armoured sub-components until late
in the manufacturing cycle. Furthermore, simplifications of the divertor geometry are presented that avoid the
need to manufacture many special cassettes, for example that might be required to accommodate the wide
range of diagnostics incorporated into the divertor. Intermediate ducts have been introduced into the design to
bridge the gap between the four pumping ports and the divertor cassettes. These ducts, in combination with gas
seals between the sidewalls of the cassettes, localize the high pressure (1-10Pa), hydrocarbon-rich exhaust gas
preventing it from entering other divertor level ports and the area behind the divertor. Apart from helping to
localize potential C-H deposits, the design reduces the overall burden of gas to be removed during the dwell
time, inhibits the exhaust gas from re-entering the main chamber during a pulse, and avoids the need to
develop elaborate gas seals around the diagnostics and viewing systems that are integrated into other ports at
the divertor level. In response to the feedback from remote maintenance specialists on the difficulties foreseen
in in-vessel handling of the FDR 2001 design of divertor cassettes, a cassette to vessel attachment scheme
aimed at simplifying the maintenance operations is proposed. Instead of clamping the cassettes to rails
attached to the vessel walls, the cassettes are maintained in position by a spring on the cassette pressing
features on the cassette into recesses in the vessel wall. The paper introduces the design, discusses the
implications for remote installation and build tolerances, and outlines the planned programme of work aimed
at validating it.
Corresponding Author:
TIVEY RICHARD
tiveyr@itereu.de
ITER Garching JWS, Boltzmannstr. 2, D-85748 Garching, germany
262
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P4C-F-92
TOWARDS THE DEVELOPMENT OF WORKABLE ACCEPTANCE
CRITERIA FOR THE DIVERTOR CFC MONOBLOCK ARMOUR.
D'AGATA ELIO, TIVEY RICHARD
Boltzmannstrasse, 2 D-85748 Garching Germany
The plasma-facing components (PFCs) of the divertor are subjected to high heat flux (HHF). Carbon-fibre
composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the
plasma intercepts the vertical targets. Failure of the heat sink to armour heat sink joints will compromise the
performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine.
There are tens of thousands of CFC to CuCrZr joints. The aim of the PFC design is to ensure that the divertor
can continue to function even with the failure of a few joints. In preparation for writing the procurement
specification for the ITER vertical target PFCs a programme of work is underway with the objective of
defining workable acceptance criteria for the PFC armour joints. This paper discusses the implications on the
operation of ITER of both the failure and the sub-standard performance of components. Based on this
understanding, acceptance criteria are proposed. Firstly, to ensure that the erosion rate and hence the carbon
released into the divertor channel is within tolerable limits, and secondly, to make it extremely unlikely that,
because of defects in the structure and/or poor thermal conductivity, the critical heat flux (CHF) will be
exceeded leading to an ingress of coolant event (ICE) into the main chamber. Most promising are the
thermographic techniques such as those developed by CEA and Plansee. These have shown that defects can be
detected in relatively thin-walled (» 5 mm) armour. However, with thick-walled armour with anisotropic
properties like that proposed for the ITER divertor, system errors, largely due to variations in thermal
conductivity, mean only relatively large defects can be detected with any certainty. This means that although
defects that might lead to an ICE can be detected, smaller defects that will contribute to a reduced armour
lifetime cannot. One solution to this problem is to use acceptance criteria that take account of the standard
deviation in the thermal conductivity of the CFC. This paper will propose such limits that should ensure a
satisfactory performance of the ITER divertor. It is important that these limits be defined in advance of any
manufacturing contract with industry, and that they are workable without reducing the demands on the
manufacturer to provide a high quality product.
Corresponding Author:
D'AGATA ELIO
dagatae@itereu.de
Boltzmannstrasse, 2 D-85748 Garching Germany
263
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P4C-F-100
RESULTS AND ANALYSIS OF HIGH HEAT FLUX TESTS ON A FULL
SCALE VERTICAL TARGET PROTOTYPE OF ITER DIVERTOR
MISSIRLIAN MARC, F. ESCOURBIAC (1) M. MEROLA (2) I. BOBIN-VASTRA(3) J. SCHLOSSER (1) A.
DUROCHER (1)
(1) CEN Cadarache, 13108 Saint-Paul-Lez-Durance (FRANCE) (2) EFDA Close Support Unit, Garching (GERMAN) (3)
FRAMATOME, Le Creusot (FRANCE)
An extensive development programme has been carried out in the EU on high heat flux components within the
ITER project. In this framework, a Full Scale Vertical Target (VTFS) prototype of ITER divertor has been
tested by means of the FE200 electron beam European facility at Framatome in Le Creusot (France). This
component was 1000 mm long and contained all the main features of the corresponding ITER design. Four
units using entirely the monoblock technology were assembled in parallel and actively water cooled. The
armour upper part of the prototype was made of an alloy of tungsten (W-1%La2O3) lamellae whereas the
lower part was made of Carbone Fibre reinforced Carbon (CFC-NB31). The heat sink was in precipitation
hardened copper (CuCrZr) equipped with a swirl insert into the straight part of the cooling channel. The
manufacturing technology was Active Metal Casting (AMCâ) followed by an Hot Isostatic Pressing (HIP)
step. The rear side of each monoblock tiles is machined in order to allow slidings in its stainless steel support
structure. Several steps of fatigue cycling on CFC and W armoured regions were planned on this prototype
taking into account ITER safety margin requirements in terms of thermal fatigue. CFC monoblocks were
tested up to 23 MW/m2 x 2000 cycles (10 s heating phase/10 s dwell phase) on the straight part without any
indication of failure. W monoblocks endured 10 MW/m2 before a first water leak after ~600 cycles and 15
MW/m2 before a second water leak after ~100 cycles. After these high heat flux experiments, metallographic
examination were undertaken on the damaged units of the prototype. This paper summarises the main test
results and describes the numerical simulation of the thermomechanical behaviour of the VTFS mock-up
during the production process as well as during the thermal fatigue loading. The purpose of the
thermomechanical analyses coupled to damage valuation is to allow a reasonable interpretation of the occurred
phenomena during this fatigue cycling campaign. Thermal and mechanical stress analyses have been
performed using the CAST-3M finite element code including transient and steady state thermal analyses as
well as fatigue life time evaluation under ITER operating conditions.
Corresponding Author:
MISSIRLIAN MARC
missir@drfc.cad.cea.fr
CEN Cadarache, 13108 Saint-Paul-Lez-Durance (FRANCE)
264
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P4C-F-115
STRUCTURAL AND FRACTURE MECHANICS ANALYSIS OF ITER
TOROIDAL FIELD COIL
SHUJI IKI, K.WATANABE(1) K.TAKIUE(2) M.SAITO(2)
(1)JFE Steel Corporation(East Japan Works),Chiba,260-0835,Japan (2)Univ.of Tsukuba,Tsukuba,Ibaraki,305-8573,Japan
One of structural issues of toroidal field (TF) coil of a tokamak fusion device is a considerably large transverse
displacement due to out-of-plane electromagnetic loads induced by the interaction of TF coil current with the
magnetic fields. The magnetic fields are varying with central solenoid (CS) coil current, poloidal field (PF)
coil current and plasma current. These currents are controlled to realize the assumed operation scenario. The
first purpose of the study is evaluations of maximum displacement and maximum stress of TF coil to confirm
the structural integrity of TF coil. The TF coils without shear panels are examined using standard operation
scenario. The lower end of a gravity support is fixed as a boundary condition and in the bonded region with
adjacent TF coil, a periodical boundary condition is employed. The numerical results are: TF coil leans toward
one direction in beginning and leans toward the reversed direction in nearly end of the operation scenario. The
maximum transverse displacement is beyond 110mm and appears in the upper curved part of TF coil. From the
viewpoint of plasma physics, the magnetic field disturbed by the TF coil displacement may be significant.
Thus in ITER, the shear panels are installed to reduce the transverse displacement. The maximum Mises stress
of about 590MPa appears in the lower bonded region in nearly end of the scenario. The lower bonded region is
loaded by a tension-compression cycle in one shot of plasma. Since the maximum stress area is restricted
locally in the structural material of TF coil case, there appears no highly stressed area inside the coil case, that
is, the superconducting material is not so strained. However, because of the stress amplitude in every shot of
plasma, the lower bonded region should be examined from the viewpoint of the possibility of fatigue crack
propagation. The second purpose of this study is the detailed evaluation of stress field around the initial crack
assumed in the lower bonded region. The stress intensity factor K is employed to investigate the crack growth.
From fracture mechanics analysis, there is no significant crack growth in the anticipated cycle of pulses in
ITER. In conclusions, globally, in a sense of structural mechanics, the electromagnetic load is large and gives a
considerably large transverse displacement of TF coil, but locally, in a sense of fracture mechanics, the
electromagnetic load is not so large and gives only a small value of K.
Corresponding Author:
SHUJI IKI
ikisyu@riko.tsukuba.ac.jp
Saito-lab,Institute of Engineering Mechanics and Systems,University of Tsukuba,Tsukuba,Ibaraki 3058573,Japan
265
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P4C-F-116
CRACK PROPAGATION BEHAVIOR AROUND DSCU/SS316 HIP
BONDED INTERFACE BY THERMAL FATIGUE
TAKAHIRO OYAMA, AKIRA YAMAMOTO(1) KENSUKE MHORI(2) MASAKATSU SAITO(1)
(1)Saito-lab, Institute of Engineering Mechanics and Systems, Univ. of Tsukuba, Tsukuba, Ibaraki, 305-8573, Japan
(2)Kawasaki Heavy Industries, Ltd., Tokyo, 136-8588, Japan
The pulse operation is assumed in ITER. If the first wall has a defect, a crack may be propagated. The first
wall is composed of DSCu and SS316. HIP bonded process is employed to fabricate the first wall. This study
deals with the crack propagation behavior around HIP bonded interface by thermal fatigue in order to confirm
integrity of HIP bonded joints. Thermal fatigue experiments were carried out by use of the EB gun as a heat
source. Specimens are DSCu/SS316 HIP bonded plates. Their thicknesses are 3.0mm and 5.0mm respectively.
A surface crack of nearly 0.5mm depth is introduced in DSCu. DSCu surface including an initial crack was
cyclically irradiated by heat flux. SS316 surface was cooled by the cooling plate. The maximum temperature
difference was about 500degree. As a result, the crack propagation direction changed perpendicularly near HIP
bonded interface. Cracks along HIP bonded interface were propagated on the surface which left the thin copper
layer (thickness about 20micro meter) to SS316. The cracks did not penetrate into SS316. The crack
propagation rate on HIP bonded interface (estimate over 1micro meter/cycle) was much larger than the crack
propagation rate in DSCu (over 0.1micro meter/cycle). 2-dimensional elasto-plastic thermal stress analysis of
DSCu/SS316HIP bonded plates was carried out by use of finite element code MARC. First, this analysis
explains the change of crack propagation direction around HIP bonded interface from the viewpoint of fracture
mechanics. Delta J hat, the amplitude of efficient J hat integral, was calculated by use of deLorenzi's virtual
crack extension method. Two delta J hat, delta J_x hat along the crack surface and delta J_y hat perpendicular
to the crack surface, were compared. When the crack tip in DSCu is far from HIP bonded interface, delta J_x
hat is larger than delta J_y hat. When the crack tip comes near to HIP bonded interface, delta J_y hat increases
rapidly. At HIP bonded interface delta J_y hat is much larger than delta J_x hat. Second, stress intensity factor
K_i defined along the interface was computed from the stress distribution. K_i decreases as the crack tip’s
highest temperature decreases.
Corresponding Author:
TAKAHIRO OYAMA
oyama@riko.tsukuba.ac.jp
Saito-lab, Institute of Engineering Mechanics and Systems, Univ. of Tsukuba, Tsukuba, Ibaraki, 305-8573,
Japan
266
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P4C-F-133
SIMULATION OF MANY-ATOMIC INTERACTIONS IN W-O-H
SYSTEM WITH THE MD CODE CADAC
I.S. LANDMAN,
In the vessel of future tokamak reactors such as ITER, tungsten is the first candidate material for the divertor
armour and perhaps for the first wall. Chemical erosion of tungsten surfaces under impact of scrape-off-layer
(SOL) deuterium-tritium plasma containing impurities, for instance oxygen, is an important issue for its use in
the reactor components. The impurities coming to the irradiated surface can form volatile molecular complexes
WxOy. The H-atoms retained near the implantation layer can create the volatile complexes HxOy, which
reduces the formation of WxOy and mitigates the corrosion effect (H states for tritium or deuterium). The
impact of the hot H-ions of the SOL plasma can destroy the complexes and influence the surface chemistry
drastically. To investigate this W-O-H system the molecular dynamics (MD) simulation code CADAC was
recently developed. In this work, the previous version of CADAC that contains a pair-atomic interaction
algorithm is extended to include many-atomic interactions. Now CADAC can simulate such important effects
as the molecular organization of atoms. To achieve this generalisation of the code, some practical however
rather general new approach that enabled to tackle complexity of the system is introduced. MD is naturally
combined with the concept of valence. The model approximates the chemical reactions using atomic valences
and available data on the atom-atom pair-wise interactions. In this work the new model itself is described for
the first time, some elementary atomic configurations built of H-, O- and W-atoms are analyzed, such
important atomic systems as O2 and H2 gases and the W-lattice with many-atomic potentials are simulated at
room temperature, and the first results on the chemical erosion of tungsten surface with the many-atomic
interactions taken into account are presented.
Corresponding Author:
I.S. LANDMAN
igor.landman@ihm.fzk.de
Forschungszentrum Karlsruhe, Institute for Pulsed Power and Microwave Technology, Post Box 3640, 76021
Karlsruhe, Germany
267
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P4C-F-135
DESIGN OF A LIMITER FOR THE JET EP ICRH ANTENNA
CHAPPUIS PHILIPPE, CHRISTOPHE PORTAFAIX(1) ERIC THOMAS(1) BERNARD BERTRAND(1) ROBERT
WALTON(2) VALERIA RICCARDO(2) RICHARD BAKER(2) IAN BARLOW(2) ALAN KAYE(2) AXEL LORENZ(3),
(1)Euratom-CEA Cadarache, CEA/DSM/DRFC, 13108 St Paul lez Durance, France (2)Euratom-UKAEA/Fusion, Culham,
OX14 3DB Abingdon, United Kingdom (3)EFDA-JET CSU-Culham, OX14 3DB Abingdon, United Kingdom
A new set of poloidal limiters have been designed and manufactured to allow the installation of the upgraded
ICRH system on the JET machine within the available wall space. The aim of the ICRH upgrade is to
demonstrate adequate power handling capability in conditions relevant to the ones to be found on ITER-FEAT
(large distance between launcher and plasma last closed flux surface and ELMs). The Antenna is surrounded
by a frame of Carbon tiles allowing for a full protection against any impinging particle flux on the straps. The
tiles are attached to an Inconel frame designed to control efficiently the flow of current between the plasma
and the vessel. The frame is connected to the vessel by mechanical supports allowing for the vessel distortions
and adjustment during installation, baking and to a lesser extent operation. The limiter beams and all the tiles
are installed in the vessel using Remote Handling (RH). The plasma facing carbon tiles were shaped to comply
with the existing limiters and with various plasma scenarios. Based on magnetic equilibrias calculated using
the finite element code Proteus, the optimisation was achieved through the CFPflux field line tracing code to
ensure the correct shadowing of all edges. Consequently the designed surface temperature should remain lower
than 900 C in the worst case. The Eddy and Halo current distribution in the limiter frame was determined with
the ANSYS using design current sinks & sources and field transients based on previous operational
experience. The total current flow is controlled by the use of resistive straps between the different limiter
elements. The mechanical calculations linked to the interaction of these currents with the magnetic field led to
the optimisation of the limiter weight (RH constrains) while maintaining allowable stresses in all elements. As
part of the ICRF Antenna project, the design of the Poloidal limiter was carried out by CEA, supervised by
EFDA CSU JET in close co-operation with the JET operator (UKAEA). After agreement by the operator
Quality Assurance (QA) system and selection of a proper qualified company by way of an international tender,
the manufacturing of all elements is ongoing and should be achieved in August 2004.
Corresponding Author:
CHAPPUIS PHILIPPE
philippe.chappuis@cea.fr
CEA DRFC, CEA Cadarache, 13108, St Paul lez durance, France
268
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P4C-F-145
EROSION OF TUNGSTEN MACROBRUSH ARMOR AFTER MULTIPLE
INTENSE TRANSIENT EVENTS IN ITER
BAZYLEV BORIS, G.JANESCHITZ (1) I.S. LANDMAN (1) A. LOARTE (2) S.E. PESTCHANYI(1)
(1) Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe, Germany (2) EFDA-CSU, Max-Planck-Institut fuer
Plasmaphysik, D-85748 Garching, Germany
In the future tokamak ITER, tungsten is foreseen as one of perspective materials for the divertor and the dome.
The main disadvantage of bulk tungsten armour is surface cracking under high heat loads typical for the
intense transient events such as disruptions and ELMs. In one ITER discharge about 1000 ELMs are expected.
During ITER operation several hundred disruptions may occur. One possibility to mitigate the surface cracking
is the tungsten macrobrush armour (W-brushes). However during the transient events, when a significant part
of confined plasma is dumped onto the macrobrush elements, it may result in surface melting of them. The
melt motion may produce significant surface roughness and droplet splashing thus causing erosion of the
elements. The separatrix strike position (SSP) at the surface can substantially vary in sequential ELMs, which
correspondingly changes the distributions of the heat flux and the pressure of impacting plasma at the target.
The results of fluid dynamics simulations for the melt motion erosion of W-brushes after multiple
stochastically varied plasma heat load pulses typical of the ITER regime are presented for the following ranges
of the surface energy deposition Q and the pulse duration t: Q = 5–100 MJ/m2 and t = 1–10 ms (disruption), Q
= 1–5 MJ/m2 and t = 0.1–0.5 ms (ELM). The heat loads are calculated applying the two-dimensional MHD
code FOREV-2D taking into account the vapour shield in front of the target and radiation transport in the
ITER magnetic field configuration for both the face- and lateral structures of W-brushes. The target melt
motion erosion is calculated by the fluid dynamics code MEMOS-1.5D. The surface tension and the viscosity
of molten metal as well as the Lorentz forces due to the currents crossing the melt layer are taken into account.
The geometric peculiarities of W-brushes and the melt motion along the gap edges as well as the features of
energy deposition in such a complicated geometry are implemented in MEMOS-1.5D remaining in frame of
the “shallow water” approximation on which the code is based. The erosion features of the tungsten bulk
armour and that of the W-brushes are compared. Latest validations of the codes against the experiments on
plasma guns and the tokamak JET are described.
Corresponding Author:
BAZYLEV BORIS
bazylev@ihm.fzk.de
IHM, Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe, Germany
269
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P4C-F-146
DEVELOPMENT OF AN ORIGINAL ACTIVE THERMOGRAPHY
METHOD ADAPTED TO ITER PLASMA FACING COMPONENTS
CONTROL
ALAIN DUROCHER, N.VIGNAL(A), F.ESCOURBIAC(A), J.L.FARJON(A) , J.SCHLOSSER(A), F.CISMONDI(B)
(a)Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 SAINT PAUL LEZ DURANCE, France (b)
Université de Toulon et du Var, BP 132 83957, LA GARDE, France
The interface inspection by active infrared thermography of actively cooled components has been a technique
used at TORE SUPRA for several year in the field of the High Heat Flux (HHF) components. An infrared
thermography test bed named SATIR (Station Acquisition Traitement InfraRouge) has been developed
specially by CEA in order to evaluate the manufacturing process quality of actively water-cooled plasma
facing components. The technical specifications for the supply of ITER Divertor Vertical Targets (DVT) stated
that all Cu cast layers on W or CFC armour shall be subjected to 100% thermographic examination, such as
the CEA developed SATIR test. Today, the current SATIR facility does not allow control for the large-sized
HHF components such as those ITER DVT. In the past, the control of full-scale ITER DVT mock-up showed
the limitations of the SATIR test bed in term of measuring accuracy, flow rate capability (2m3/h), low
pressurization of water loop (3bar), and useful hot water volume (1.2m3). However the last studies realised on
SATIR allowed to define the features of a new installation named "SATIRPACA" adapted to the control of
ITER DVT: - Obviously, the principle of the SATIR test bed by internal thermal excitation remains a very
interesting method and must be preserved because it measures exactly ability of the component to be cooled.
Moreover in 2004 important improvements have been performing about the detection sensitivity. - An unique
Non Destructive Examination method at this level of high technology is not sufficient. The original coupling
of SATIR with a lock-in thermography system will be realized in this study, which will allow to have two
infrared thermographic inspection methods on the same test bed. An improvement of the global reliability of
the coupled facility is expected by merging the data produced by the two techniques. Afterwards the using of
the cooling channel of HHF component during the test will improve the sensitivity of detection by the lock-in
thermography method. - The heat transfer convective coefficient will be also improved by installing an
upgraded flow rate device. The new proposed enhanced test bed is designed for the full scale Non Destructive
Examination of the HHF ITER components. The control of Wendelstein-7X HHF components expected from
the beginning 2005 will allow to validate SATIRPACA test bed. This paper proposes a detail overview of
improvements which will equip this new infrared test bed.
Corresponding Author:
ALAIN DUROCHER
alain.durocher@cea.fr
Association Euratom-CEA, CEA/DSM/DRFC,CEA/Cadarache, F-13108 SAINT PAUL LEZ DURANCE, France
270
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P4C-F-162
PLASMA SPRAYED TUNGSTEN-BASED COATINGS AND THEIR
PERFORMANCE UNDER FUSION RELEVANT CONDITIONS
MATEJICEK, JIRI, VLADIMIR WEINZETTL (1) YOSHIE KOZA (2)
(1) Institute of Plasma Physics, Za Slovankou 3, 18221 Praha, Czech Republic (2) Forschungszentrum Juelich, IWV-2, D52425 Juelich, Germany
Tungsten is one of the candidate materials for plasma facing components for ITER and other fusion devices.
Plasma spraying is among prospective fabrication technologies, thanks to its ability to coat large areas and the
possibility of in-situ repair. This paper reports on the development of tungsten and tungsten+copper plasma
sprayed coatings and their behavior under high heat fluxes and in tokamak plasma. Several tungsten-based
coatings were produced at IPP, using water-stabilized plasma spraying from different powders and under
various spraying conditions. Their basic properties (structure, composition, etc.) were characterized by SEM,
XRD and other techniques. The coatings contain varying amount of porosity and oxides; these factors are
subject to further process optimization. The behavior of plasma sprayed coatings and a solid tungsten sample
under high temperature plasma conditions was investigated at the small-size CASTOR tokamak at IPP (ohmic
heating 30 kW, pulse length of 30 ms, electron temperature ~30-60 eV, ion temperature ~10-20 eV, plasma
density 0.5-1x10-19 m-3). The samples were inserted into plasma at various radii and exposed to standard
plasma discharge conditions. Main plasma parameters were observed, together with impurity radiation in the
XUV, VUV and visible ranges. Only a local influence of the tungsten presence was found. Heating and
cooling rates were measured between the shots. No melting and very limited surface modification of the
materials was observed. Selected coatings were tested under high heat fluxes at the electron beam facility
JUDITH at FZJ, to simulate disruption conditions. The samples were subjected to different incident beam
current (30-125 kV) and loading time (5-10 ms) over the area of ~7 mm2, while the absorbed current, surface
temperature and particle emission were recorded. The induced changes were observed by surface profilometry,
SEM and optical microscopy. These were namely the removal of oxide scale at lower incident energies,
surface melting at intermediate and deep melting at high energies. The coatings were able to absorb about 0.5
GW/m2 (2.5 MJ/m2) in thermal shock loading without significant damage.
Corresponding Author:
MATEJICEK, JIRI
jmatejic@ipp.cas.cz
Institute of Plasma Physics, Za Slovankou 3, 18221 Praha, Czech Republic
271
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P4C-F-167
HIGH TEMPERATURE STRESSES IN ITER RELEVANT BRAZED
GLIDCOP/W MODEL STRUCTURES
COPPOLA* ROBERTO, C. NARDI 1, M. VALLI 2
(1) ENEA-Frascati, FUS, CP 2400, 00100 Roma, Italy (2) ENEA-“Clementel“, V. Don Fiammelli 2, 40129 Bologna, Italy
It is well known that plasma-facing components of near term machines, such as ITER, must be designed to
withstand huge and long (1000 s) heat charges with consequent thermo-mechanical stresses. Therefore, the
knowledge of the stress field in components such as the divertor is essential to define the engineering
parameters required to design real-scale components. This contribution will present the results of an
experimental study on high temperature stress evolution in brazed Glidcop/W model structures. The samples,
approximately 23 x 23 x 8 cm3 in volume, were obtained by brazing a W and a Glidcop platelet at 650 C using
TiCu Ag alloy as a filler. Neutron diffraction was utilized to determine the strains, then the stresses in the bulk
of the samples between room-temperature and 500 C; unstrained Glidcop and W reference samples were
measured as well. The measurements were carried out at the D1A diffractometer available at the High Flux
Reactor of the ILL-Grenoble. The experimental results provide relevant engineering information such as the
zero strain temperature. The adopted experimental procedure will also be discussed in view of its possible use
for real-scale components.
Corresponding Author:
COPPOLA* ROBERTO
coppolar@casaccia.enea.it
ENEA-Casaccia, FIS, CP 2400, 00100 Roma - I
272
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P4C-F-176
A MATURE INDUSTRIAL SOLUTION FOR ITER DIVERTOR PLASMA
FACING COMPONENTS: HYPERVAPOTRON COOLING CONCEPT
ADAPTED TO TORE SUPRA FLAT TILE TECHNOLOGY
ESCOURBIAC FREDERIC, V.KUZNETSOV(2) M.MISSIRLIAN(1) B.SCHEDLER(3) J.SCHLOSSER(1)
(1) Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 SAINT PAUL LEZ DURANCE, France (2)
Efremov institute, Doroga na Metallostroy, St. Petersburg, 196641, Russia (3) Plansee AG, 6600 Reutte, Austria
The design heat flux for specific plasma facing components in ITER is in the same range (10-20 MW/m²) than
observed in electron tubes. Historically, concepts with enhanced cooling capabilities implying
boiling/condensation effects due to a fin design named hypervapotron were developed by Thomson CSF tube
Company for such purposes and later designed for neutral beam heating systems. This cooling concept adapted
to a CuCrZr heat sink armoured with CFC or W was envisaged for the vertical targets of the ITER divertor
since the beginning of ITER EDA, but finally abandoned for two main reasons : it was suspected that the joint
temperature between CFC or W and CuCrZr may be too high as well as a possible occurrences of a “cascade
tile failure” effect. Last experimental results accompanied with progress in modelling have shown excellent
behaviour of hypervapotron based armours with regard to the two mentioned supposed disadvantageous
arguments : temperature of the armour/heat sink joint - strongly dependent on the flow velocity – can be driven
below a tolerated limit of 500 C and cascade tile failure occurrence was not experimentally observed. In order
to validate the hypervapotron concept as a design solution for the ITER divertor, thermal fatigue testing has
been performed on two medium scale mock-ups. They were manufactured by Plansee AG with respect to the
main technological features of a TORE SUPRA toroidal limiter finger element. One of these mock-ups was
tested in the European facility FE200 (Elecron Beam 200 kW) and the other one in the Russian facility
TSEFEY-M (Elecron Beam 60 kW). Both testing campaigns have shown that the mock-ups were able to
sustain with margins corresponding to the divertor requirements in terms of thermal fatigue : 3000 cycles at 15
MW/m², 800 cycles at 25 MW/m² and a critical heat flux limit higher than 30 MW/m². Analyses of tests
results will be reported in this paper. (*)High heat flux testing partially supported by EFDA
Corresponding Author:
ESCOURBIAC FREDERIC
frederic.escourbiac@cea.fr
(1) Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 SAINT PAUL LEZ DURANCE,
France
273
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P4C-F-192
CONCEPTUAL DESIGN OD A HIGH-TEMPERATURE WATERCOOLED DIVERTOR FOR A FUSION POWER REACTOR
GIANCARLI LUCIANO, J.P. BONAL (1), A. LI PUMA (1), B. MICHEL (2), J.F. SALAVY (1), P. SARDAIN (3)
(1) CEA/Saclay, DEN/DM2S, 91191 Gif-sur-Yvette, France (2) CEA/Cadarache, DEN/DER, 13108 St.Paul-lez-Durance,
France (3) EFDA, CSU-Garching, Boltzmannstr. 2, D-85740 Garching, Germany
The large effort devoted in recent years to the divertor development for ITER reactor has shown that the
divertor is a critical reactor component because of the severe operating conditions which have to be withstand,
such as very high surface heat fluxes, interaction with energetic plasma particles, and complex geometry.
Additional requirements need to be fulfilled for a power reactor divertor such as the resistance to high neutron
fluxes and fluences, use of high temperature coolant for achieving an acceptable overall reactor thermal
efficiency, and, possibly, the use of low activation materials. This paper presents the studies performed in the
framework of the EU Power Plant Conceptual Study (PPCS) concerning the development of the conceptual
design of a water-cooled divertor using low-activation martensitic steel (EUROFER) as structural material,
water coolant at PWR conditions (15.5 MPa pressure and 325 C outlet temperature), and W-alloy monoblock
as armour. The concept consists of a series of EUROFER pipes for coolant flow, each of them surrounded by a
W-alloy monoblock, attached to a common EUROFER back plate. The concept is able to withstand a
continuous surface heat flux of 15 MW/m2, reaching an acceptable maximum structure temperature of about
516 C and showing acceptable stresses, provided an appropriate interface between pipes and monoblock is
used. EUROFER has been selected due to its expected capability of withstanding neutron damages higher than
70 dpa (eq. Fe). However, because of its relatively low thermal conductivity and its differential thermal
expansion with the W-alloy, direct joints EUROFER/W-alloy cannot be used in order to avoid too high
thermal stresses. Therefore it is proposed to add at the interface a thermal barrier on the front half of the pipes,
made of pyrolitic graphite to enhance the thermal flux repartition, and of a compliance layer made of soft
graphite “papyex”. In this case, thermal stresses become acceptable and the maximum W-alloy temperature is
about 2000 C. The main issues of this divertor concept are the manufacturing process of the steel/W interface
and the behaviour under irradiation of graphite materials. Experimental data up to 30 dpa (eq. C) have been
collected in the literature and their assessment, presented in this paper, shows that the behaviour of such
materials, when used as thin layers without mechanical functions, could be acceptable.
Corresponding Author:
GIANCARLI LUCIANO
luciano.giancarli@cea.fr
CEA/Saclay, DEN/CPT, 91191 Gif-sur-Yvette, France
274
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P4C-F-195
DEVELOPMENT OF A COPPER ALLOY TO BERYLLIUM HIP
BONDING TECHNOLOGY FOR THE ITER FIRST WALL
P. SHERLOCK (1), A. T. PEACOCK (2) A. D. MC CALLUM (1)
(1) NNC Limited, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England (2) EFDA CSU Garching,
Boltzmannstr. 2, D-85748 Garching, Germany
The Primary First Wall (PFW) modules of the ITER blanket concept are covered with separable PFW panels.
The PFW panels comprise a bi-metallic copper alloy / stainless steel 316L water-cooled heatsink faced with a
plasma facing material. Dispersion Strengthened copper (DS-Cu) and precipitation strengthened CuCrZr are
options for the copper alloy. One option for the plasma facing material is beryllium, in the form of tiles. Over
recent years, the technology needed to bond beryllium tiles to the copper alloy of the heatsink has been
developed. During this development, solid HIP bonding has been employed as one method to produce the
heatsink base and bond the beryllium tiles in place. The development of the manufacture is typically done in
three stages. Small samples are first produced in the laboratory to show that beryllium can be bonded to
CuCrZr under ideal conditions using the selected parameters. Larger mock-ups are then produced which have
some geometrical aspects that are similar to those of the full size panels. This shows the selected bonding
parameters can be used under manufacturing conditions. The final stage is to produce full size prototype PFW
panels to prove the technology at this scale [1]. During the first two stages structural analysis, mainly in the
form of finite element analysis, has been used to assess the mechanical behaviour during the manufacturing
process. For the small samples, the analysis models the stresses that result from the differential thermal
expansion between the beryllium and copper alloy. This assists in the selection of the compliant layer which
strains to accommodate the expansion and reduce residual stress. On the larger mock-ups, the interaction
between the HIP can and the mock-up itself during the HIP processes is modelled to progress the design of the
HIP can / mock-up assembly. This paper describes the small samples and larger mock-ups produced by NNC
during the development of the copper alloy / beryllium HIP bonding technology. It demonstrates how
structural analyses were used to gain an understanding of the bonding process and develop the HIP can /
mock-up assembly design. [1] Manufacture of blanket shield modules for ITER, P. Lorenzetto et. al., this
conference.
Corresponding Author:
P. SHERLOCK (1)
paul.sherlock@nnc.co.uk
NNC Limited, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England
275
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P4C-F-228
AN ADVANCED HE-COOLED DIVERTOR CONCEPT: DESIGN,
COOLING TECHNOLOGY, AND THERMOHYDRAULIC ANALYSES
WITH CFD
IHLI, THOMAS (1), R. KRUESSMANN (1), I. OVCHINNIKOV (2), P. NORAJITRA (1), V. KUZNETSOV (2), R.
GINIYATULIN (2)
(1) Institute for Materials Research III, Forschungszentrum Karlsruhe GmbH, P.O. Box 3640, 76021 Karlsruhe, Germany (2)
D.V. Efremov Institute, Scientific Technical Center "Sintez", 196641 St. Petersburg, Russia
An advanced modular helium-cooled divertor concept for near-term reactor models like DEMO is being
investigated at Forschungszentrum Karlsruhe (FZK). It is based on the multiple jet impingement cooling
technology which is efficiently applied in the gas turbine sector. The major challenge for the divertor concept
is to handle target heat loads of up to 10-15 MW/m². Hot helium is chosen as coolant due to its advantageous
safety characteristics. It allows for a high exit temperature of at least 700 to 750 C, which is suitable for the
power conversion system that uses a gas turbine cycle. A highly effective and reliable cooling system is
necessary to fulfil the requirements. Nevertheless, the pumping power for the coolant should be kept as low as
possible and the thermal stresses caused by extreme temperature differences in the heat-loaded and cooled
parts have to be kept below acceptable limits. This leads to a segmented bodywork for the targets which
consist of small multiple finger units (e.g. 9 fingers in their own housing). The fingers consist of small Wamour parts, that are brazed onto separate pressure-carrying components (caps). The interior of the caps is
cooled by arrays of helium jets supplied from impingement hole arrays in cartridges which are inserted into the
caps. The finger units can be tested separately before fixing them to stripe units which form the targets. The
divertor system presented is extremely flexible and can be adapted to all kinds of gas-cooled reactors by
adjusting the jet hole configuration and the numbers of parallel and series connections of the small multiple
finger units. It ensures a high heat transfer coefficient at a well-balanced mass flow rate. The performance of
the concept is investigated by means of finite-element (FE) and computational fluid dynamics (CFD) analyses.
The results of a CFD parameter study, focusing on the minimum helium mass flow rate required for cooling
and the respective temperature distributions are incorporated in the thermohydraulic design. They provide for
an iterative approach comprising CFD and FE calculations for design improvement and the prediction of
pressure loss and heat transfer coefficients prior to experimental investigations of the concepts. In this study,
the design and cooling method shall be described briefly. Design performance and layout examples shall be
highlighted using results of CFD calculations performed at FZK and the Efremov Institute.
Corresponding Author:
IHLI, THOMAS (1)
thomas.ihli@imf.fzk.de
Institute for Materials Research III, Forschungszentrum Karlsruhe GmbH, P.O. Box 3640, 76021 Karlsruhe,
Germany
276
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P4C-F-239
THE NEW ELECTRON BEAM TEST FACILITY JUDITH II FOR HIGH
HEAT FLUX EXPERIMENTS ON PLASMA FACING COMPONENTS.
MAJERUS PATRICK, REINER DUWE (2) TAKESHI HIRAI (1) WINFRIED KÜHNLEIN (2) JOCHEN LINKE (1)
MANFRED RÖDIG (2)
(1) IWV 2, Forschungszentrum Jülich GmbH, EURATOM-Association, 52425 Jülich, Germany (2) B-Z, Forschungszentrum
Jülich GmbH, EURATOM-Association, 52425 Jülich, Germany
The Juelich Divertor Test Facility in Hot Cells JUDITH I has been operating successfully in the Research
Centre Jülich since the early nineties. It represents a unique high heat flux experiment for simulating thermal
loads on neutron activated plasma facing components. The potential for testing toxic materials, like beryllium
and for simulating all ITER relevant thermal loads (including disruptions, VDEs and ELMs) made JUDITH I
break its capacity during the recent years. To extend the parameter range and because of the urgent need of
additional testing capacity a new electron beam facility JUDITH II is being build up. Beyond the ability to
perform the same type of experiments as in JUDITH I, a whole range of optimised features is included into the
new facility. An approximately three times higher nominal power of 200 kW, combined with a beam scanning
angle of ±14 enables to test larger components up to 0,5 x 1 m². The relatively small acceleration voltage,
adjustable between 30 and 60 kV, reduces volumetric heating for the benefit of a more plasma like surface
heating. Due to a very flexible and individual programmable system for electron beam pattern generation, a
highly homogeneous load distribution can be achieved. Two different testing modes, pulsed and beam sweeper
mode, allow rather realistic simulations of the ITER relevant transient loads. Power densities up to 10 GW/m²
and minimum event times as short as 2 µs are the only limiting parameters. This offers a full range of new
possibilities in simulating ELMs with deposited energy densities in the order of 1 MJ/m² and pulse durations
of several hundred microseconds. ELMs have only recently been identified as possibly life-time limiting
events in future confinement experiments, such as ITER. Furthermore it will become possible to combine
static and transient loads in one single experiment. Besides IR-diagnostics, high resolution and fast image
grabbing will allow an enhanced study of the effects caused by transient loads on armour candidate materials.
Especially brittle destruction will be addressed, using a spectrometer in the visible range, a photodiode array
and acoustic emission. The first two methods serve to analyse the emitted particle while acoustic emission
shall give information on the onset of brittle destruction. With a combination of the applied methods it is
expected to additionally measure the energy release rate per emitted particle.
Corresponding Author:
MAJERUS PATRICK
p.majerus@fz-juelich.de
IWV-2, Forschungszentrum Jülich GmbH, EURATOM-Association, 52425 Jülich, Germany
277
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P4C-F-253
FORMATION OF CRYSTALLINE NANOSTRUCTURES DURING
DEUTERIUM PLASMA INTERACTION WITH TUNGSTEN-BASED
MATERIALS IN SIMULATED GAS DIVERTOR CONDITIONS.
GUSEVA MARIYA, V.M. GUREEV, L.S. DANELYAN, B.N. KOLBASOV, S.N. KORSHUNOV, V.B. PETROV, B.I.
KHRIPUNOV
Nuclear Fusion Institute, RRC Kurchatov Institute, Kurchatov sq. 1, 123182 Moscow, Russia
Sputtering of W-based materials considered as candidates for ITER divertor armour manufacturing [W–10Re,
W–1La2O3, W–13I and W(111)] by deuterons with subthreshold energy (5 eV) in a dense steady-state plasma
of the LENTA facility was studied by weighing technique at armour temperatures from 1250 to 1520 K. It has
been found that sputtering of these materials occurs above 1250 K. At 1520 K and irradiation dose of
1.5x10*26 m*-2, the sputtering yields for W(111), W–13I, W–10Re and W–1La2O3 are 1.1x10*-4, 2.6x10*4, 2.9x10*-4 and 5.3x10*-4 respectively. Specimen weight loss decreases with increase in irradiation dose, e.g.
the weight loss of the W–10Re specimen reduced fivefold when the irradiation dose doubled (to 3x10*26 m*2). Microstructure studies and X-ray diffraction analysis suggest that such an effect is due to the formation of
some special structures on the W surfaces. At an irradiation dose of 3x10*26 m*-2, surfaces of the specimens
acquire a block nanostructure consisting of various polygons, including pentagons, hexagons and heptagons of
different areas. The smallest blocks (~100-nm) were observed on the W–13I surface and the largest (0.1-3
mm) – on the W–1La2O3 surface. The presence of pentagons, hexagons and heptagons may be attributed to
the condensation of sputtered atoms and the formation of a new type of substance under irradiation by
deuteron flux of 10*18 cm*-2s*-1. X-ray diffraction analysis suggests that the W structures on the specimen
surfaces are strongly textured in the <110> direction. The W lattice spacing (3.165 Å) is the same for all the
specimens within the limits of experimental error. The structures on the surfaces of W–10Re, W–13I and
W(111) are practically continuous and have a good adhesion. Those covering W–1La2O3 (PLANSEE) have
1–2-mm holes of irregular shapes. Deuterium (D) and protium (H) were detected in a narrow, ~20 nm thick,
near-surface film layer using elastic recoil detection analysis. D content in this layer is insignificant (<0.05
at.%). H concentration is 2.5 at.%. Thus, our findings suggest that surface structures with negligible D content,
preventing further erosion of W-based ITER divertor elements would arise on the surface of these elements
even after 7-8 normal (400-s) pulses, provided the elements’ temperature is kept in the range of 1250-1520 K.
It would be of interest to look into the possibility of using the nanostructures forming on the W surfaces for
technological purposes.
Corresponding Author:
GUSEVA MARIYA
sinet@nfi.kiae.ru
Nuclear Fusion Institute, RRC Kurchatov Institute, Kurchatov sq. 1, 123182 Moscow, Russia
278
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P4C-F-265
ACTIVITY OF THE EUROPEAN HIGH HEAT FLUX TEST FACILITY:
FE200
I. BOBIN-VASTRA, F. ESCOURBIAC (2), M. MEROLA (3), P. LORENZETTO (3)
2) CEA-DRFC-SIPP, Cadarache, 13108 St Paul lez Durance (F) 3) EFDA Close Support Unit, Boltzmannstr. 2, D-85748
Garching, Germany
FE200 is an Electron beam (EB) 200KW test facility stemming from partnership between Framatome
Technical Center in Le Creusot (F), and Tore Supra team in CEA Cadarache (Euratom/CEA association),
dedicated to high heat flux testing of plasma facing components for fusion devices. Since 1992, an extensive
development program has been carried out in the FE200 high heat flux facility especially for the ITER and
TORE SUPRA tokamaks. In this framework, more than 100 000 cycles for thermal fatigue tests, 400 critical
heat fluxes for different hydraulic conditions, 200 disruptions and 2 tests with glancing incidence (cascade
failure) were performed on materials such as Cu-Al25 and CuCrZr alloys, Carbon Fibre Composite (CFC) or
Tungsten (W) monoblocks and tiles, and plasma spray-W coatings for various actively cooled plasma facing
component designs. The tests concerned small mock-ups as well as large components with a length ranging
from a few tens of millimeters up to 1m in length. The facility includes a 200KW EB gun (200KV, 1A) which
is able to deliver continuously during more than one hour from 0.1 to >100 MW/m² heat flux for thermal
fatigue testing and up to 10GJ/m² during a few milliseconds for disruptions. A programmable sweeping allows
several kinds of energy repartition (uniform and peaked), on a 13 shooting angle. In the 8m3 vacuum
chamber, the maximum allowable length for components to be tested, is 1m if the surface is perpendicular to
the beam, 2m when the component is tilted. The component is connected to a pressurised loop working from
0.2 to 3.3 MPa, at temperatures between 50 to 230 C, up to a maximum flow rate of 6kg/s. This large range of
parameters gives a flexibility to the pressurised loop, which allows LOFA tests (Loss of Flow Accident) with
successive cooling rate modifications during the test. Instrumentation gives information for diagnostics on
calorimetry balance (absorbed heat flux), surface temperature till 2300 C (Infrared camera and pyrometers
with remote positioning during test), visual aspect or behaviour (CCD camera with remote focusing). The
paper illustrates the FE200 capabilities through several testing scenarios on different mock-ups and
components tested in this facility, namely cascade failure configuration results on CFC and W, thermal fatigue
tests performed on Primary First Wall (PFW) type components and divertor component, highest critical heat
fluxes and LOFA tests on hypervapotron designed mock-ups.
Corresponding Author:
I. BOBIN-VASTRA
isabelle.bobinvastra@framatome-anp.com
Framatome-anp Centre Technique (groupe AREVA), FE200, Porte Magenta, BP181, 71205 Le Creusot Cedex
(F)
279
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P4C-F-274
PROPOSAL OF LITIZATION OF FTU VACUUM VESSEL BY USING A
LITHIUM LIMITER
APICELLA MARIA LAURA, G. MAZZITELLI (1) V.B. LAZAREV (2) E.A. AZIZOV (2) S.V. MIRNOV (2) V.G.
PETROV (2) V.A. EVTIKHIN (3) I.E. LYUBLINSKI (3) A.V. VERTKOV (3) F. LUCCA (4)
(1) Ass. ENEA-EURATOM sulla Fusione CR Frascati (2) Troitsk Inst. for Innovation and Fusion Research, Troitsk, Moscow
Reg., RF (3) State Enterprise «Red Star» - Prana-Center Co, Moscow, RF (4) L.T. Calcoli SaS, Via C. Baslini, 13 - 23807
Merate (LC)
A new promising idea for the application of liquid lithium as plasma facing material in fusion reactors has
been recently proposed and began to be tested. It is based on the surface tension forces in capillary channels
that may be used to compensate forces induced in liquid lithium by the JxB effect under the plasma MHD
events in tokamaks. The new structure, called CPS (Capillary Porous System) has been realized as a matt from
wire meshes of Stainless Steel 304 with pore average radius 15 micron and wire diameter 30 micron. The
liquid lithium flows inside these capillaries from on side of the system, which is in contact with a liquid
lithium reservoir, to the other side that is faced to the plasma. The main features of CPS are the high stability
and resistance to surface damage and the self-regeneration of the lithium surface through capillary forces. This
last property becomes very important for divertor plates and the wall protection of ITER-like or post-ITER
tokamak that will operate in the presence of ELMs which are the main reason of enhanced erosion. FTU, a
medium size tokamak, represents a very good opportunity to test for the first time CPS configuration for a
litization experiment by using a liquid lithium limiter. This experiment consists in the wall coating with a thin
lithium film produced during a plasma discharge by a displacement of the LCMS (Last Closed Magnetic
Surface) towards the lithium limiter. Before its installation on FTU, foreseen for the beginning of 2005, a
detailed study of plasma scenario and lithium limiter operations have been done. In addition, a full
electromagnetic analysis of the JxB forces and their influence on liquid lithium confinement in the capillaryporous limiter under disruption in FTU (Ip=1.6 MA, BT = 8 T) has been carried out. These results indicate
that, assuming a realistic shape of CPS (three modules nearly semi cylindrical) and a thickness of CPS layer
equal to 1mm, the peak amplitude of electromagnetic pressure can reach the value of 10 kPa which is a factor
5 lower than the capillary pressure able to retain liquid lithium in porous structure. The study of the lithium
limiter experiment on FTU has been completed with a thermal and thermal-mechanical analysis performed by
ANSYS 5.2 code.
Corresponding Author:
APICELLA MARIA LAURA
apicella@frascati.enea.it
ENEA C.R. Frascati - Via E. Fermi, 45-00044 Frascati-Roma-Italia
280
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P4C-F-278
DESIGN, PERFORMANCE AND CONSTRUCTION OF A 2 MW ION
BEAM TEST FACILITY FOR PLASMA FACING COMPONENTS
GREUNER, HENRI, B. BOESWIRTH, T. FRANKE, P. MCNEELY, N. RUST
A new ion beam test facility for the testing of plasma facing components (PFCs) under high heat fluxes is
presently under construction at IPP Garching. The aim of this facility is to provide thermal testing capabilities
for high heat loaded PFCs with both active water cooling and large outer dimensions. Start of operation is
planned to be late summer 2004. Long pulse and cycling heat load tests of WENDELSTEIN 7-X divertor
target elements and of complete target modules are the main activities planned for the next years to ensure the
successful development, manufacturing and operation of these components. The experience gained from these
extensive tests can be used to later adapt the facility to the requirements of effective HHF testing for ITER
divertor components. The facility consists of a water-cooled vacuum vessel with a diameter of 1.5 m, a length
of 3.7 m and is equipped with 2 ion sources. Initially, only one of the two individually controlled RF ion
sources with 1.1 MW maximum beam power will be used for heat load tests in an operating regime between 5
and approximately 65 MW/m² at the target position. The water-cooled ion source allows for high power, long
pulse operation facilitating cycling tests of large components. The water cooling system of the facility is
designed for testing of components with cooling water consumption of up to 8 l/s and a pressure drop of 15
bar. This paper describes the technical characteristics and operating conditions of the facility. The vacuumand cooling system, the power supply and control system, the target diagnostic and data acquisition system are
described in detail.
Corresponding Author:
GREUNER, HENRI
henri.greuner@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, D- 85748 Garching, Germany
281
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P4C-F-279
SPECTROSCOPIC STUDIES OF HOMOGENEOUS CARBON FLAKES
WITH A HIGH DEUTERIUM CONTENT FORMED IN TOKAMAK T-10
STANKEVICH VLADIMIR, N.YU. SVECHNIKOV (1), A.M. LEBEDEV (1), K.A. MENSHIKOV (1), B.N.
KOLBASOV (1), L.N. KHIMCHENKO (1), N.M. KOCHERGINSKY (2), D. RAJARATHNAM (2), YU. KOSTETSKI
(2)
(1) Russian Research Center “Kurchatov Institute”, Kurchatov sq. 1, 123182 Moscow, Russia (2) Faculty of Engineering,
National University of Singapore, Singapore
Carbon films redepositing on plasma facing elements in tokamaks attract the attention of investigators mainly
as accumulators of hydrogen isotopes, especially tritium. Homogeneous deuterated carbon a-C:D films with a
high deuterium content [1], redeposited under deuterium plasma discharges inside the T-10 tokamak vacuum
chamber, have been studied using thermogravimetric analysis (TGA), electron spin resonance (ESR) , Fouriertransform infrared (IR) reflection, as well as luminescence spectroscopy in vacuum ultraviolet and visible light
ranges, including luminescence excitation by synchrotron radiation in the range 4-18 eV at 300 K. TGA
measurements have revealed that a mass loss of up to 30% occurs at 450 C mainly at the expense of carbon
and water. ESR spectroscopy results point to a high density of free radicals (~10*19 spins) and a low
anisotropy g-factor. As for IR spectroscopy results, we found a significant decrease of deformational
vibrations of CHx aromatic groups (out of plane and in plane) in the wavelength range below 1000 cm-1 after
baking at 450 C. It indicates that high amount of aromatic groups, including protium was desorbed during
baking. On another hand, C-H and C-D stretching modes have shown different behaviour: the peak
corresponding to sp3 C-H stretching mode (2925 cm-1) increased after baking at 450 C, possibly due to the
decay of OH stretching modes with a subsequent H hopping to C-radicals, while the peak of C-D stretching
mode slightly decreased. The photoluminescence effect, observed at 390-530 nm for the first time, could be
related to C2p pi-pi transitions within aromatic rings with a subsequent electron-hole recombination. The
luminescence was quenched at 450 C – apparently due to the formation of disordered aromatic network within
the gap states. The latter possibility is supported by a high value of spins formed on the in-gap defect states.
The luminescence excitation spectra of the tokamak films are similar to those of fullerene C60 films for peaks
at 3.4, 6.5 and 8.5 eV of a C2p pi- and sigma- character, which appeared to be common for tokamak a-C:D
films and C60 systems. [1]. P.V. Romanov, B.N. Kolbasov, V.Kh. Alimov, et al. J. Nucl. Mater. 307-311
(2002) 1294.
Corresponding Author:
STANKEVICH VLADIMIR
VGS@polyn.kiae.su
Russian Research Center “Kurchatov Institute”, Kurchatov sq. 1, 123182 Moscow, Russia
282
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P4C-F-280
VACUUM PLASMA-SPRAYED TUNGSTEN ON EUROFER AND 316L RESULTS OF CHARACTERISATION AND THERMAL LOADING
TESTS BOLT, HARALD, H. GREUNER, B. BOESWIRTH, S. LINDIG, W. KÜHNLEIN (1), T. HUBER (2), K. SATO (3), S.
SUZUKI (3)
(1) FZ Jülich, Euratom Association, Forschungszentrum Jülich, B-NZ Heisse Zellen, 52428 Jülich, Germany (2) PLANSEE
AG, A-6600 Reutte/ Tirol, Austria (3) Blanket Engineering Laboratory, JAERI, Naka-machi, Naka-gun, Ibaraki-ken, 3110193 JAPAN
Tungsten is being considered as a potential plasma facing material for future fusion devices, primarily due to
its low erosion rate and heat resistance. Vacuum plasma spraying (VPS) of tungsten is an effective industrial
technique for coating actively cooled plasma facing components made of low activation steels or stainless steel
316L. The coated material would be a potential candidate for first wall components receiving moderate heat
load up to 1MW/m². A development programme examined the manufacturing and suitability of W-VPS
coatings as plasma facing material on up to 1 MW/m² heat loaded first wall components. Mock-ups made of
martensitic steels EUROFER and F82H as well as austenitic steel 316L were coated with 2 mm thick W-VPS
layers. Mixed tungsten/steel interlayers were applied to both reduce the residual and thermal stresses at the
substrate-coating interface and to improve the adhesion of the coating,. The characterisation of the W-VPS
layers included the evaluation of the coating micro structure, the measurement of physical and mechanical
properties (thermal conductivity, density, hardness, bending strength and Young’s modulus etc.) and the
metallographical examination before and after heat load tests. Thermal loading tests were carried out at the
JUDITH facility at the Research Centre Jülich and in parallel in the JEBIS facility at JAERI for the F82H
mock-up. Successfully completed cycling tests with heat loads of 2 MW/m² and screening tests up to 2.5
MW/m² confirm the thermomechanical suitability of W-VPS coatings for plasma facing first wall components
made of steel.
Corresponding Author:
BOLT, HARALD
harald.bolt@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, D- 85748 Garching, Germany
283
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P4C-F-285
CAN TOKAMAK DEVICES SURVIVE ELMS DURING NORMAL
OPERATION? A SIMULATION STUDY
KONKASHBAEV, ISAK, HASSANEIN, AHMED
Argonne National Laboratory, 9700 S. Cass Ave., Bldg. 308, Argonne, IL 60439, USA
During normal operation of H-mode, edge-localized modes (ELMs) are serious concern for divertor and
nearby plasma-facing components (PFCs) of the next generation tokamaks. During ELMs part of the total
plasma energy is released and deposited on divertor surface in duration 0.1-1 ms with a frequency of 1-20 Hz
depending on ELM type. The power from scrape-off-layer (SOL) to PFC in ITER-like devices can then
increase from 5 MW/m2 to ¡Ö 300-3000 MW/m2. Erosion lifetime strongly depend on ELM power deposited.
However, the resulting evaporated material can reach the core and disrupt the plasma. In addition, with higher
ELMs frequency, thermal cycling takes place and can result in thermal stresses and fatigue. At high ELM
power, the resulting high surface temperature causes vapor cloud formation with similar consequences to
disruptions. Vapor shielding decreases energy deposition at the surface but increases radiation flux to nearby
components. Metallic PFC can melt and liquid metal flow instabilities occur with mass losses due to both
MHD splashing effects and vaporization. In this study a comprehensive two-fluid model is developed to
integrate Core and SOL parameters during ELMs with PFC surface evolution (melting, vaporization, vapor
cloud hydrodynamics and mixing with plasma particles, and macroscopic spallation) for low and high ELM
power using HEIGHTS numerical simulation package. Initial results indicate that high-power, i.e., Giant
ELMs in ITER-like machines can cause serious damage to PFCs, may terminate plasma in disruptions, and
because of large contamination may affect subsequent plasma operations. A comparison of modeling results
with available data from current machines is also addressed.
Corresponding Author:
KONKASHBAEV, ISAK
hassanein@anl.gov
Argonne National Laboratory, 9700 S. Cass Ave., Bldg. 308, Argonne, IL 60439, USA
284
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P4C-F-294
EU R&D ON DIVERTOR COMPONENTS
MEROLA MARIO, W. DAENNER (1) M. PICK (1)
(1) EFDA, Boltzmannstr. 2, D-85748 Garching, Germany
Selected also for oral presentation
O4A-F-294
Since the last SOFT conference held in Helsinki in 2002, substantial progress has been made in the EU R&D
on the divertor components. A number of activities have been completed and new ones have been launched.
The present paper gives an update of the works carried out by the EU Participating Team in support of the
development of the divertor, which is one of the most challenging components of the next step ITER machine.
One of the most impressive achievement was the further development and consolidation of suitable
technologies for the production of high heat flux components with both CFC and tungsten armour joined onto
a copper alloy heat sink, namely CuCrZr. This long lasting effort culminated with the manufacturing of a near
full-scale vertical target prototype, which was high heat flux tested well above the ITER design loads. To
ensure competition among the EU industries, different technologies were developed like HIP’ing, brazing and
Hot Radial Pressing. Work is now focused on the preparation of EU industry to the unprecedented ITER series
production. Another recent achievement was the completion of the post-irradiation testing of divertor mockups and material samples. This activity demonstrated that the proposed technologies are able to perform above
the requirements, even after being neutron irradiated at 0.2 and 1.0 dpa at 200 C. A substantial design effort
was also carried out in collaboration with the ITER IT, with the EU Associations and EU industries. The final
outcome is a comprehensive ITER divertor design capable to withstand all the expected loads with minimum
manufacturing costs, minimum waste and maximum performances. The manufacturing of a complete set of
full-scale divertor prototypes with dummy armour was launched and is progressing according to schedule.
After their delivery, the PFCs will be used to validate a software tool, which was recently specifically
developed by an EU Association to simulate the hydraulics of the ITER divertor including the draining and
drying. In preparation for writing the procurement specification for the ITER vertical target PFCs, an activity
is in progress in the EU with the objective of defining workable acceptance criteria for the PFC armour joints.
It will be the experimental basis for the final definition of the maximum acceptable defects as well as to assess
if and how these defects can be detected by means of non-destructive testing techniques.
Corresponding Author:
MEROLA MARIO
mario.merola@tech.efda.org
EFDA, Boltzmannstr. 2, D-85748 Garching, Germany
285
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P4C-F-305
THERMAL MODELING OF W ROD ARMOR SUBJECTED TO ELMS
NYGREN, RICHARD E.,
Sandia National Laboratories* has been developing and testing mockups armored with tungsten (W) rods for
most of the last decade and Sandia pioneered the initial development of W rod armor for ITER in the 1990's.
Plasma facing components (PFCs) with W rod armor have been designed for the ITER-FEAT divertor and are
the reference design for the FIRE divertor. Water-cooled heat sinks armored with tungsten rod armor can
endure heat fluxes near 25 MW/m2 without cracking, melting or debonding of the armor. Results from earlier
have been reported at SOFT and elsewhere and we continue the testing program. We have also developed 3-D
thermal models of the W rod-armored PFCs and applied the model to both short pulse testing to simulate ITER
ELMs (edge localized modes) and thermal performance in steady state. The basic model is 1/6 of an individual
rod with mirror boundaries along the cut sides. Variations include (1) a model with cells graduated to 10
microns in thickness at the top of the rod to follow the shallow and rapid thermal penetration of very high heat
laods (1-5 GW/m2) necessary to model the high energy density from ELMS, (2) a rod based upon a W rod
cluster to be tested in the DiMES probe in DIII-D, and (3) aggregates of several rods to study the effects of
uneven heating on rod groups. The mesh was created in PATRAN and the model is run using ABAQUS 6.3.1.
Heat loss due to thermal radiation is included as are temperature dependent properties of materials. The heat
transfer coefficient at the water cooling boundary follows a specified boiling curve that depends on the coolant
temperature, pressure and the presence of any heat transfer enhancement such as a twisted tape insert. This
paper briefly describes the model and focuses on the thermal modeling of ELMs. For example, the threshold
energy density for melting was studied for various values of steady state heat flux (i.e., starting surface
temperature). Also, in the case of repeated ELMs that caused some melting, the model accounts for the
enthalpy stored in melting and its effect on the thermal response during ELMS repeated with frequencies of 110 Hz. Applications of the model to a W rod DiMES probe to various heat loads and the thermal performance
of W-rod-armored mockups are also mentioned. *Sandia is a multi-program laboratory operated by Sandia
Corporation, a Lockheed Martin Company, for the United States Department of Energy under Contract DEAC04-94AL85000
Corresponding Author:
NYGREN, RICHARD E.
renygre@sandia.gov
Sandia National Laboratories MS1129, PO Box 5800, Albuquerque NM 87185 USA
286
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P4C-F-315
CRITICAL HEAT FLUX TESTING ON SCREW COOLING TUBE MADE
OF RAFM-STEEL F82H FOR DIVERTOR APPLICATION
KOICHIRO EZATO, SATOSHI SUZUKI, MASAYUKI DAIRAKU, KAZUYOSHI SATO, AND MASATO AKIBA
As part of development of Plasma-Facing Components (PFCs) for fusion machines, JAERI has been
developing high performance cooling tubes with pressurized water flow. Along this line, a cooling tube with a
helical triangular fin on its inner surface has been proposed recently for application to a DEMO reactor. Since
the fin can be machined by a simple mechanical threading, this tube is called as a screw tube. In our previous
experiments, it was reported that heat removal performance of the screw tube made of pure Cu is twice as high
as that of a smooth tube. In DEMO designs, Reduced Activation Ferritic Martensitic (RAFM) steel such as
F82H is one of the candidate materials for a cooling structure of PFCs instead of Cu-alloy. As thermal
conductivity of F82H is about ten times smaller than that of Cu-alloy, this study is intended to examine heat
removal capability of the screw tube made of F82H. For this purpose, we have carried out Critical Heat Flux
(CHF) testing under one-sided heating conditions by using a hydrogen ion beam. The test samples are the
screw tubes with M10 of 1.5-mm-pitch. The M10 threads are directly shaped in F82H and OFHC-Cu tubes
with the outer diameter of 12 mm. The minimum wall thickness of each tube is 1 mm. Inlet temperature and
local pressure of cooling water are room temperature and 1MPa. Flow velocity ranges from 2 to 12 m/s.
Incident heat flux at the sample position has a Gaussian profile and its maximum value ranges from 8 to
48MW/m2. Incident CHF (ICHF) of the F82H screw tube is reduced to about half of the OFHC-Cu tube. For
instance, ICHF of the F82H tube is 13 MW/m2 at the flow velocity of 4 m/s and that of the Cu tube is
25MW/m2. Numerical analyses show that the critical heat flux at the inner surface of the cooling tube is
almost the same for both tube materials. This means that the incident heat flux is highly concentrated for the
F82H cooling tube because of its low thermal conductivity. It is also found that the ratios of the heat flux at the
inner surface of the cooling tube to the incident heat flux are around 1.6 for the F82H tube and 1.1 for the Cu
tube. Based on these results it turns out that application of F82H to PFC cooling structures needs to enhance
dispersion of the incident heat flow, for example, to be covered with armor material with higher heat
conductivity such as tungsten.
Corresponding Author:
KOICHIRO EZATO
ezatok@fusion.naka.jaeri.go.jp
Japan Atomic Energy Research Institue, 801-1 Mukoyama, Naka-machi, Naka-gun, Ibaraki-ken 311-0193,
Japan
287
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P4C-F-328
STATUS OF HE-COOLED DIVERTOR DEVELOPMENT FOR DEMO
NORAJITRA , PRACHAI, GINIYATULIN, RADMIR (B) IHLI, THOMAS (A) KRAUSS, WOLFGANG
(A) KRUESSMANN, REGINA (A) KUZNETSOV, VLADIMIR (B) MAZUL, IGOR (B) OVCHINNIKOV,
IVAN (B)
(a) FORSCHUNGSZENTRUM KARLSRUHE, P.O. BOX 3640, D-76021 KARLSRUHE, GERMANY (b) D.V. Efremov
Institute, Scientific Technical Centre “Sintez”, 196641 St. Petersburg, Russia
The helium-cooled divertor is considered a suitable option for fusion power plants, as it is compatible with
likewise He-cooled blanket systems. It is also recommended for those blankets, where water-cooling of invessel components would lead to considerable concerns in terms of safety (e.g. steam-beryllium reaction with
H production). Furthermore, it allows for a relatively high gas outlet temperature, i.e. a high thermal efficiency
of the power conversion systems. He-cooled modular divertor concepts with integrated flow promoters in the
form of a pin (HEMP) or slot (HEMS) array are being developed at the Forschungszentrum Karlsruhe within
the European Power Plant Conceptual Study. In parallel, an alternative design HEMJ is under investigation,
which is based on multiple jet impingement cooling without flow promoter. The modular design helps to
reduce thermal stresses. Tungsten is considered the most promising material to withstand the high heat load,
due to its high melting point, high thermal conductivity, and low thermal expansion. The proposed HEMP/S
divertor concept employs small W tiles of quadratic or hexagonal shape, which are brazed to a thimble
structure of W alloy below. A flow promoter of W is brazed underneath each thimble to increase the cooling
surface. The structure is made of high-temperature ODS RAFM. The development and optimisation of the
divertor concepts require a close link of and iterative approach comprising the main issues of design, analyses,
materials, fabrication technology, and experiments. Predicting the temperatures and stresses by means of
computational fluid dynamics and finite element computer codes is indispensable to ensure that the
engineering design limits are not exceeded. The divertor working temperature window is restricted by ductilebrittle transition temperature at the lower and recrystallisation temperature of the W structure at the upper
limit. Enlarging this temperature window is a challenging task of materials development. For manufacturing
divertor components of W and W alloy, EDM, ECM, laser, and PIM are considered promising methods.
Experiments on W/W and W/steel joining were performed successfully at Efremov. A helium loop will be
built at Efremov this year for high-heat-flux (HHF) integral testing of the divertor design variants. An
electronic beam facility there allows for the HHF simulation of 10 MW/m² at least. The status of development
in the above areas of work shall be outlined in this report.
Corresponding Author:
NORAJITRA , PRACHAI
norajitra@imf.fzk.de
FORSCHUNGSZENTRUM KARLSRUHE, P.O. BOX 3640, D-76021 KARLSRUHE, GERMANY
288
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P4C-F-333
NUMERICAL AND EXPERIMENTAL STUDY OF DEMO HE-COOLED
DIVERTOR TARGET MOCK-UPS
RUMYANTSEV MIKHAIL, KUZNETSOV VLADIMIR (1) OVCHINNIKOV IVAN (1) FILATOV VLADIMIR (1)
JANESCHITZ GUENTER (2)
(1) The D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, 3 Doroga na Metallostroy, Promzona
"Metallostroy", St. Petersburg 196641, Russian Federation (2) Forschungszentrum Karlsruhe, P.O Box 3640, D 76021
Karlsruhe, Germany
The helium cooled divertor with modular conception is proposed for the fusion reactor DEMO. It is planned to
use helium (600oC, 10MPa) for divertor cooling, where a high surface heat flux (up to 15MW/m2) must be
removed from vertical targets. For the best cooling of modules the smooth surfaces which interacts with
helium flow are increased with help of additional pins or slots. The more area of the surface the better cooling,
but at the same time pumping power may be increased, i.e. useful power will be decreased. Selection of an
optimal geometry of the cooling surface is current problem of project. Optimization of cooling surface at
expected reactor conditions was performed with help numerical code. The following options of modules are
presented: - a first option is modules with straight radial slots on cooling surface. An optimization of height,
width and numbers of slots was performed for this option. Influence of increasing in target dimensions and
helium mass flow rate on thermal state were analyzed here, too. By results of these analyses a best geometry of
slots was selected. Thermal stresses were examined for this geometry. - a second geometry option is module
with pins. Also paper describes experimental gas puffing facility GPF2, where the analyzed mock-ups were
tested. Reversed heat flux was proposed for the test facility which consists from 2 loops, helium (600oC, 10
MPa), and water (RT, 5 MPa). In this case, the mock-up with the hot helium flow is intensively cooled by
water from the plasma-facing side. This approach give possibility to study: different mock-up designs can be
compared with respect to pressure drop and cooling efficiency. The GPF2 works with helium pulses that are
longer by 2-3 orders of magnitude to reach stationary flows in the mock-up. The simulation approach, methods
and data processing are described. Test results obtained for different mock-ups at fluxes of 5-15 MW/m2 are
presented and discussed. In addition, behavior of the mock-ups in GPF2 was simulated by CFD and the
calculation results are compared with the experiments.
Corresponding Author:
RUMYANTSEV MIKHAIL
rumyanmi@sintez.niiefa.spb.su
Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA) 3 Doroga na Metallostroy,
Promzona "Metallostroy", Metallostroy, St. Petersburg 196641
289
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P4C-F-343
MEASUREMENTS OF H/D DIFFUSIVITY IN AND SOLUBILITY
THROUGH TUNGSTEN IN THE TEMPERATURE RANGE OF 600 C TO
800 C
AIELLO ANTONIO, GIANLUCA BENAMATI (1) ANDREA CIAMPICHETTI (2)
(1) ENEA C.R. Brasimone - Bacino del Brasimone, 40032 Camugnano (BO)- Italy (2) Politecnico di Torino – DENER –
Corso duca degli Abruzzi 24, 10129 Torino - Italy
Estimation of tritium permeation through the plasma facing materials is an important issue for fusion reactors.
Because of their refractory nature and good thermal properties, tungsten and tungsten-alloys are considered to
be alternatives to graphite as plasma-facing materials for ITER. Tungsten has a very high threshold for
sputtering as well as a high melting point. Tungsten is expected to be used in areas where the energy of plasma
particles can be kept well below the sputtering threshold, removing the plasma impurities problem associated
with the use of this material. Experimental campaigns on the characterisation of hydrogen isotopes transport
and solubility parameters of tungsten have been conducted by several laboratories, but results are often in
disagreement. An extensive experiment have been conducted in ENEA using a permeation device named PERI
2. Permeation tests were carried out using membranes of tungsten separating two volumes in the PERI II
apparatus, an high pressure volume and an high vacuum volume. Hydrogen gas was charged in the high
pressure side and measuring the pressure evolution in the low pressure side it was possible to determine the
hydrogen transport parameters in the sample. A permeated gas analysis by means of a mass quadrupole was
also performed. Experiments conducted in the temperature range between 350 C and 800 C indicated the low
permeability of tungsten. The obtained results together with the experimental procedure adopted are herein
presented and discussed.
Corresponding Author:
AIELLO ANTONIO
antonio.aiello@brasimone.enea.it
ENEA C.R. Brasimone - Bacino del Brasimone - 40032 Camugnano (BO) Italy
290
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P4C-F-367
TESTING OF ACTIVELY COOLED MOCK-UPS IN SEVERAL HIGH
HEAT FLUX FACILITIES – AN INTERNATIONAL ROUND ROBIN
TEST
ROEDIG, MANFRED, I. BOBIN-VASTRA (2), S. COX (3), F. ESCOURBIAC (4), A. GERVASH (5), A.
KAPOUSTINA (1), W. KUEHNLEIN (1), V. KUZNETSOV (5), M. MEROLA (6), R. NYGREN (7), D.L.
YOUCHISON (7)
(1) FZJ, Jülich, Germany (2) Framatome, Le Creusot, France (3) JET, Abingdon, UK (4) CEA-DRFC, Cadarache, France
(5) Efremov Inst., St. Petersburg, Russia (6) EFDA Close Support Unit, Garching, Germany (7) Sandia Nat. Lab.
Albuquerque, USA
In next step fusion devices like ITER, the first wall and divertor component will be exposed to high heat fluxes
up to more than 10MW/m2. Hence a large R&D effort is being carried out to develop suitable high heat flux
components. In order to test components under operational relevant conditions, several electron or ion beam
facilities have been used worldwide. Up to a certain degree these machines are comparable. They consist of a
beam generator, a beam sweeping system, a vacuum test chamber, and a number of diagnostic devices. But
some machine parameters like the beam generation and system, calibration techniques and diagnostic systems
are quite different. In order to assess the influence of these differences on testing results, a round robin test has
been performed on five electron beam facilities a few years ago. The aim was, to study the influence of
specific layouts of these machines on the results of high heat flux tests. The comparison was carried out by
high heat flux testing of actively cooled CFC samples at identical target loading conditions, and the surface
temperatures were used as a criterion for the assessment of the results. This former test campaign was not
planned as a round robin test from the very beginning, and some questions stayed open. Hence a new test
campaign has been initiated by the EFDA team. For this new test campaign special actively-cooled mock-ups
have been produced, and testing parameters have been planned with respect to the testing facilities involved. In
the beginning, only electron beam facilities were intended to take part. But later the program was extended,
and tests at the JET neutral beam injector testbed have been included. In this testing campaign, a set of actively
cooled CFC monoblock mock-ups has been loaded in the different facilities at comparable power densities.
The temperature response during these loadings on the surface (IR cameras, pyrometers) and inside the mockups (thermo couples) has been registered and used as a criteria for comparison. Furthermore finite element
calculations have been carried out for the temperature fields at different power densities. Most of the surface
temperatures were found in a relatively narrow scatter band. Only one of the electron beam facilities shows
somewhat higher temperatures compared to the other machines. At higher power densities, the JET-NBI data
are on the upper side of the scatter band. This may be explained by the peaked beam profile in this machine.
Corresponding Author:
ROEDIG, MANFRED
m.roedig@fz-juelich.de
Forschungszentrum Juelich, 52425 Juelich, Germany
291
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P4C-F-376
STUDY OF TECHNOLOGICAL AND MATERIAL ASPECTS OF HECOOLED DIVERTOR FOR DEMO REACTOR
GERVASH ALEXANDER, R.GINIYATULIN 1, W.KRAUSS 2, A.MAKHANKOV1, I. MAZUL 1, P.NORAJITRA 2
1 Efremov Research Institute, 196641 St. Petersburg, Russia 2 Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021
Karlsruhe, Germany
Study of technological and materials aspects of He-cooled divertor for DEMO reactor Further development of
a helium-cooled divertor concept for fusion reactor like DEMO depends strongly on the progress in selection
of suitable materials and technologies of their manufacturing and joining as well. Design proposes small
tungsten tiles joined to a thimble structure made of tungsten alloy. High-temperature ferritic steel is proposed
for supported structure. Presented paper gives recent results of technological trials to manufacture heliumcooled divertor module. To enhance inner thimble surface with pin and slot array several fabrication
techniques were checked. In particularly, the numbers of prototypes were produced by electric discharge
machining (EDM), electrochemical milling (ECM), laser ablation and chemical vapour deposition (CVD)
methods. The accuracy of required dimensions, cost reasons, possibilities of serial production of such
prototypes were compared and discussed. To select most suitable W-alloy for the thimble production paper
gives the main results of comparative testing of candidates (W-single crystal, W-1%La2O3, CVD-tungsten,
W-Cu composite, forged/rolled sintered tungsten). Investigating the problem of joining W-thimble to ferritic
steel structure the number of W/ferritic steel specimens produced by e-beam welding, diffusion bonding, high
temperature brazing and locking with cast copper were manufactured and tested. The main results are
presented and discussed. Summarizing presented data authors inform about nearest further steps of their
investigation.
Corresponding Author:
GERVASH ALEXANDER
gervash@sintez.niiefa.spb.su
Efremov Research Instutute, 196641 Saint Petersburg, Russia
292
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P4C-F-386
DESIGN AND THERMAL PERFORMANCE OF SURFACE-BOLTLESS
MECHANICALLY ATTACHED MODULE FOR DIVERTOR PLATE OF
LHD
KUBOTA YUSUKE, MASUZAKI SUGURU(1), MORISAKI TOMOHIRO(1), TOKUNAGA KAZUTOSHI(2), AND
NODA NOBUAKI(1)
(1)National Institute for Fusion Science, Oroshicho 322-6, Toki 509-5292, Japan (2)Research Institute for Applied
Mechanics, Kyushu Univ.,Kasugai 816-8580, Japan
Abstract: To prevent plasma collapse during plasma confinement experiment, especially in steady state
operations, suppressions of outgassing and high Z impurity emission from the first walls are required strongly.
According to the requirements, high performance mechanically attached module has been developed as the
next type of divertor plate for large helical device(LHD). The most advantage of the module is not to have any
bolts for fix on the armor tile surface to avoid high Z impurity emission different from the previous one1) used
in the LHD since the third campaign(FY 1999). The new one consists of two armor tiles made of iso-graphite,
a thin super graphite sheet, and a SS cooling pipe. A couple of armor tiles sandwiches the cooling pipe through
a super graphite sheet to improve thermal contact between two materials, and fixed tightly with only two
TZM(an alloy of molybdenum) bolts. This simple structure without a copper heat sink allows smooth heat
flow from the tile surface to the cooling pipe different from the previous one. Using a test facility ACT2) with
a 100kW electron gun, steady high heat flux tests up to 1.2 MW/m2 were carried out for the new one without
any trouble although the previous one was limited about 0.3 MW/m2. Moreover, outgassing from the new one
during high heat flux tests up to 0.5 MW/m2 decreased to about one-third of that of the previous one. Low
outgassing of the new one may originate in the simple structure without a copper heat sink and use of super
graphite sheet with an excellent thermal conductivity. Thermal fatigue test up to 500 cycles under steady heat
flux of 1 MW/m2 for the new one is scheduled to carry out using ACT before full-scale application to divertor
plates of LHD. The design, thermal performance, thermal analysis by 3D CAD, and outgassing of the newly
developed mechanically attached module are presented. References 1) N.Noda, S.Sakamoto, Y.Kubota et al.,
J.Plasma Fusion Res. SERIES, Vol.3(2000)180. 2) Y.Kubota, N.Noda, A.Sagara et al., Fusion Engineering
and Design 56-57(2001)205.
Corresponding Author:
KUBOTA YUSUKE
kubota@LHD.nifs.ac.jp
National Institute for Fusion Science, Oroshicho 322-6, Toki 509-5292, Japan
293
- F - Plasma Facing Components.
P4C-F-413
MANUFACTURE OF BLANKET SHIELD MODULES FOR ITER
LORENZETTO PATRICK, BOIREAU B. (2), BOUDOT C. (2), BUCCI PH. (3) FURMANEK A. (1), IOKI K. (4),
LIIMATAINEN J. (5), PEACOCK A. (1), SHERLOCK P. (6), TÄHTINEN S. (7)
(1) EFDA CSU Garching, Germany. (2) FRAMATOME ANP, Le Creusot, France. (3 ) CEA, Grenoble, France. (4) ITER IT,
Garching, Germany. (5) Metso Powdermet, Tampere, Finland. (6) NNC Ltd, Knutsford, England. (7) VTT Industrial Systems,
Espoo, Finland.
Selected also for oral presentation
O4A-F-413
The ITER Blanket-shield concept is a modular configuration mechanically attached onto the vacuum vessel
and consists of Limiter modules and Primary First Wall (PFW) / Shield modules. The latter consist of a watercooled 316L(N)-IG Stainless Steel (SS) Shield Block and separable PFW panels mechanically attached onto
the Shield Block. The PFW panels consist of a bi-metallic structure with a 316L(N)-IG SS backing plate and a
Copper (Cu) alloy heat sink layer. There are two Cu alloy candidates: Dispersion Strengthened Cu-Al25 and
Precipitation Hardened CuCrZr alloys. Beryllium (Be) tiles are joined to the Cu alloy heat sink as plasma
facing material. A Research and Development programme for the ITER Blanket-shield has been implemented
in Europe to provide input for the design and the manufacture of the full-scale production components. It
involves in particular the fabrication and testing of mock-ups and scale-one prototypes of Shields and PFW
panels. These prototypes aim at demonstrating the fabricability of the components. Two methods are being
considered in Europe for the manufacture of the Shield blocks. The first method is based on conventional
fabrication techniques using drilling, machining and welding. The second method uses a more advanced
technique based on Hot Isostatic Pressing (HIPping) of 316L(N)-IG SS powder and 316L(N)-IG SS solid
parts. One Shield prototype made from powder HIPping is already complete and a second prototype made
from mixed powder and solid HIPping is under fabrication. Two methods are also being considered in Europe
for the manufacture of the bi-metallic structure of the PFW panels: solid and powder HIPping. With solid
HIPping, the 316L(N)-IG SS backing plate, the Cu alloy plates and the 316L(N)-IG SS tubes are joined
together with one single HIP cycle. With powder HIPping, a first HIP cycle is used to consolidate the
316L(N)-IG SS powder with embedded 316L(N)-IG SS tubes. A second HIP cycle is then performed for
consolidating and joining the CuCrZr powder. Beryllium tiles are then joined by HIPping or brazing; high
temperature HIPping or furnace brazing for PFW panels with CuAl25 alloy heat sink material, and low
temperature HIPping or inductive brazing for PFW panels with CuCrZr alloy heat sink material. Three panel
prototypes have already been completed. Two more are under fabrication. This paper describes the main
fabrication steps for the above Shield and PFW panel prototypes.
Corresponding Author:
LORENZETTO PATRICK
patrick.lorenzetto@tech.efda.org
EFDA CSU Garching, Boltzmannst. 2, D-85748 Garching, Germany
294
- F - Plasma Facing Components.
P4C-F-426
THE MAST IMPROVED DIVERTOR
DARKE ANDREW, R J HAYWARD G F COUNSELL K HAWKINS
The Mega Amp Spherical Tokamak (MAST) at Culham is one of the leading world machines studying the
spherical tokamak (ST) concept. At the time of the initial construction in 1998 little was known about the sort
of divertor structures that would be required in an ST. The machine was therefore provided with relatively
rudimentary structures that were designed mostly to protect important components from the hot plasma. While
these have served the machine well it was accepted that they might not be suitable when operating MAST to
its full potential. The years of experience of operating MAST have led to the design, manufacture and now
installation of a new divertor, the MAST Improved Divertor or MID, that should be able to cope with the full
performance of the machine. The design is based on imbricated (fan-shaped) rings of tiles at the top and
bottom of the machine for the outer strike points, giving an excellent compromise between power handling and
diagnostic access, with substantial new centre column strike point armour and a shaped plate in between. High
purity graphite is chosen as the plasma facing material in preference to CFC since in this case it has a better
balance of performance and cost. The lower imbricated ring is insulated in alternate sectors for studies of
divertor biasing and extensive diagnostics and additional inboard gas injection are included. This work was
funded jointly by the United Kingdom Engineering and Physical Sciences Research Council and by
EURATOM.
Corresponding Author:
DARKE ANDREW
andrew.darke@ukaea.org.uk
EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB, UK
295
- F - Plasma Facing Components.
P4C-F-429
OXYGEN IMPURITY EFFECTS ON HYDROGEN ISOTOPE RELEASE
FROM PLASMA CHEMICAL VAPOR DEPOSITION BORON COATING
MAKOTO OYAIDZU (1), MAKOTO OYAIDZU(1), AKIRA YOSHIKAWA(1), YOSHIHIRO ONISHI(1), HHIROMI
KIMURA(1), YASUHISA OYA(2), MASAO MATSUYAMA(3), AKIO SAGARA(4), NOBUAKI NODA(4), AND KENJI
OKUNO(1)
(2)RI Center, The University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-0032, Japan (3)HIRC, Toyama University,
Gofuku 3190, Toyama 930-8555, Japan (4)NIFS, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan
The reduction of impurities, particularly oxygen, in plasma has been one of the key issues for the large
tokamaks, since plasma impurities cause dilution of fuel particles. To reduce oxygen impurities in plasma,
boronization has been developed. As a result of boronization, oxygen impurities have been dramatically
trapped in boron coating and boron coatings contained oxygen are formed on the PFM. During boronization
and/or PCVD processes, hydrogen originated from borane gases is mixed with the boron coating. Moreover, in
D-T fusion reactors, hydrogen isotopes, deuterium and tritium, are implanted into boron coatings. Therefore it
is important to elucidate hydrogen isotopes release behavior from boron coating contained oxygen from the
viewpoint of tritium retention, hydrogen recycling, and characterization. In the present study, oxygen impurity
effects on hydrogen isotopes release behavior from boron coating contained oxygen prepared using PCVD
technique was studied by XPS and TDS. Boron coatings contained oxygen deposited on silicon substrates
using a decaborane gas (vol. 20%) diluted with helium (Vol. 50%) and oxygen (vol. 30%) gases by PCVD
were used as samples. The atomic composition ratio in samples was estimated by XPS. TDS technique was
also applied to evaluate release behavior of hydrogen isotopes, namely hydrogen mixed in the samples during
preparation and deuterium implanted into the sample. XPS measurements showed that the atomic ratio of
boron to oxygen is almost unity. After TDS measurement after preparation, hydrogen release spectrum was
found to consist of a shoulder at around 400 K and a peak at around 530 K. In the hydrogen release from pure
boron coating, the temperature of low and high temperature side is around 450 and 650 K, respectively [1]. As
the result of numerical analysis, it was found that the amount of hydrogen retention in boron coating contained
oxygen is almost five times as much as that in the pure boron coating and the hydrogen release spectrum was
divided into two peaks, namely peaks A and B. Therefore, it is suggested that oxygen existing in boron coating
lead to increase hydrogen retention and hydrogen release temperature region wholly shifts to low temperature
side. In the presentation, oxygen impurities effects on hydrogen isotopes release behavior from boron coating
will be discussed in detail, taking into account for results of deuterium release behavior. [1] H. Kodama, et. al.,
J. Nucl. Mater., in press.
Corresponding Author:
MAKOTO OYAIDZU (1)
r5444008@ipc.shizuoka.ac.jp
(1)RRL, Faculty of Science, Shizuoka University, 836 Oya, Shizuoka 422-8529, Japan
296
- F - Plasma Facing Components.
P4C-F-431
IMPLANTATION TEMPERATURE DEPENDENCE ON DEUTERIUM
BEHAVIOR IN HIGHLY ORIENTED PYROLITIC GRAPHITE
HIROMI KIMURA, YASUTOMI MORIMOTO HIROSHI KODAMA MAKOTO OYAIDZU AKIRA YOSHIKAWA
TSUYOSHI TAKEDA KENJI OKUNO
Radiochemistry Research Laboratory, Faculty of Science, Shizuoka University, Ohya, Shizuoka 422-8529, Japan
For evaluation of tritium safety in fusion reactors, it is very important to know how tritium implanted into
plasma facing materials (PFMs) behaves chemically. In PFMs, tritium chemically interacts with damages
induced by energetic atom and/or irradiation. Chemical interactions are governed by two processes, thermal
processes, such as thermal release of deuterium and the thermal annealing of graphite structure, and high
energy process. In our previous study using TDS and XPS techniques, recovery of graphite structure damaged
by ion implantation was observed above 573 K. In addition, it was found that deuterium implantation at lower
temperature than RT is very useful to investigate those processes. In present study, dynamics of deuterium
implanted into graphite in various temperatures is studied from the kinetic point of view, using TDS and XPS
techniques. The sample used in the present study was a HOPG crystal purchased from Pechiney Co. Ltd.
Deuterium ions were implanted into HOPG with an energy of 1.0 keV D2+, a flux of 1.0×1018 D+ m-2 s-1,
and a fluence of 6.4×1021 D+ m-2 at various temperatures region from 173 to 773 K. To investigate
implantation temperature dependence on deuterium retention, the sample after implantation was heated up to
~1400 K with heating rate of 0.5 K s-1. To estimate activation energy of deuterium desorption, the sample was
heated with heating rate in the range from 0.083 to 1.0 K s-1. From the results of TDS experiments after
deuterium ion implantation from 173 to 773 K, deuterium implanted into HOPG was released from ~800 K.
However, the deuterium retention was decreased as implantation temperature increased. The activation
energies in the higher temperature than 237 K were determined to be ~2.3 eV, which were approximately
coincident with the literature values [1, 2], while the activation energy at 173 K was estimated to be 4.4 eV. It
was consider that this value also would include movement of carbon. In present study, taking into account for
carbon mobilization in HOPG, dynamics of deuterium implanted into graphite will be discussed in detail. [1]
T. Tanabe, et. al., J. Nucl. Mater., 179-181 (1991) 231-234. [2] K. Ashida, et. al., J. Nucl. Mater., 128-129
(1984) 792-797.
Corresponding Author:
HIROMI KIMURA
r0332004@ipc.shizuoka.ac.jp
Radiochemistry Research Laboratory, Faculty of Science, Shizuoka University, Ohya, Shizuoka 422-8529, Japan
297
- F - Plasma Facing Components.
P4C-F-445
MANUFACTURING AND TESTING IN REACTOR RELEVANT
CONDITIONS OF BRAZED PLASMA FACING COMPONENTS OF THE
ITER DIVERTOR
GRATTAROLA MARCO, M. BISIO (1) V. BRANCA (1) M. DI MARCO (2) A. FEDERICI (1) G. GUALCO (1) P.
GUARNONE (1) U. LUCONI (2) M. MEROLA (3) C. OZZANO (1) G. PASQUALE (2) S. RIZZO (1) F. VARONE
(1)
(1) Ansaldo Ricerche s.r.l., C.so Perrone 25, I-16161 Genova, Italy (2) FN s.p.a. SS 35 bis dei Giovi km 15, I-15062 Bosco
Marengo (AL), Italy (3) EFDA CSU Garching,Boltzmannstr. 2, D-85748 Garching, Germany
Selected also for oral presentation
O4A-F-445
A fabrication route based on brazing technology has been developed for the realization of the High Heat Flux
Components for the ITER Vertical Target and Dome-Liner. The divertor vertical target is based on a
monoblock design with CfC and tungsten armour in the lower straight part and in the upper curved part,
respectively. The cooling tubes are made of precipitation hardened copper alloy CuCrZr. A pure copper
interlayer between the heat sink and the armour mitigates the joint interface stress due to the thermal expansion
mismatch between the CuCrZr and the armour material. The plasma facing units of the Dome component are
based on a tungsten flat tile design with hypervapotron cooling. The heat sink is taken from a bimetallic plate
composed of precipitation hardened copper alloy CuCrZr and stainless steel joined together by means of the
explosion bonding process. An innovative brazing technique based on the addition of carbon fibers to the
active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production
(patent pending), has been used for the CFC/Cu joint to reduce residual stresses. The tungsten-copper joint has
been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required
mechanical properties of the CuCrZr alloy after brazing. The yield stress, the ultimate tensile strength and the
average grain size measured after the brazing thermal cycle are 263 MPa, 404 MPa and 28 microns
respectively. The CuCrZr/steel explosion bonding has been qualified by means of an extensive metallurgical
and mechanical test program. Non destructive examination methods based on ultrasonic techniques have been
developed and qualified to inspect the brazed joints. The fabrication route of plasma facing components for the
ITER Vertical Target and Dome based on the brazing technology has been proved by means of thermal fatigue
tests performed on mock-ups in reactor relevant conditions. Flat tile and monoblock mock-ups with both CFC
and tungsten armour material have been successfully tested by thermal fatigue tests with incident heat flux
higher than 15 MW/m2.
Corresponding Author:
GRATTAROLA MARCO
marco.grattarola@ari.ansaldo.it
Ansaldo Ricerche s.r.l., C.so Perrone 25, 16161 Genova, Italy
298
- F - Plasma Facing Components.
P4C-F-447
DEVELOPMENT OF THE PLASMA FACING COMPONENTS FOR THE
DOME-LINER COMPONENT OF THE ITER DIVERTOR
LUCONI UMBERTO, V. BRANCA (1) E. CORDANO (1) M. DI MARCO (2) A. FEDERICI (1) M. GRATTAROLA
(1) M. MEROLA (3) C. OZZANO (1)
(1) Ansaldo Ricerche s.r.l., C.so Perrone 25, I-16161 Genova, Italy (2) FN s.p.a. SS 35 bis dei Giovi km 15, I-15062 Bosco
Marengo (AL), Italy (3) EFDA CSU Garching,Boltzmannstr. 2, D-85748 Garching, Germany
On the basis of the design and the specification of the Dome-Liner elaborated by EFDA, a manufacturing route
has been developed and proved by means of the fabrication and testing of several samples and mock-ups. The
dome is supported on four posts and is protected with a 10-mm thick tungsten tile armour. These tungsten tiles
are joined onto a plate (heat sink) made of precipitation hardened copper-chromium-zirconium alloy (CuCrZr).
Due to constraints posed by the electromagnetic loads, the thickness of the CuCrZr plate can not exceed 10
mm; as a consequence the remaining part of the heat sink shall be made of stainless steel. Therefore the heat
sink is obtained from a CuCrZr/AISI 316 L bimetallic plate realized by explosion bonding. The hypervapotron
cooling channels are obtained by machining the plate from the steel side crossing the CuCrZr/AISI 316 L
explosion bonded joint, then the channels are closed by welding a steel rear closure plate. Both the explosion
bonded joint and the rear plate welding must comply the strict ITER vacuum tightness requirements. The
brazed joint between the tungsten tiles and the CuCrZr heat sink has been qualified by means of thermal
fatigue tests on small-scale mock-ups performed by the Efremov Institute (St. Petersburg, Russia) in reactor
relevant conditions (1000 cycles at 15 MW/m2). The CuCrZr/AISI 316L explosion bonding process utilized to
join the front CuCrZr plate to the rear steel backing has been qualified by means of an extensive metallurgical
and mechanical test program according to the specification provided by EFDA. The test program included the
execution of hot pressure helium leak tests, cyclic pressure and burst tests on several relevant mock-ups. The
mock-ups sustained He leak tests at 250 C and 6 MPa (pressurized He inside the components), before and
after the repeated pressure test at 8 MPa for 100 cycles. After the previous tests, the mock-ups sustained an
internal pressure higher than 500 bar during the bust tests. The dimensional stability during the fabrication
route has been investigated by means of the realization of a relevant curved component that has been
dimensionally tested after the completion of each step of the manufacturing route. The results of the
experimental activity are presented and discussed in this paper.
Corresponding Author:
LUCONI UMBERTO
umberto.luconi@ari.ansaldo.it
Ansaldo Ricerche s.r.l., C.so Perrone 25, 16161 Genova, Italy
299
- F - Plasma Facing Components.
P4C-F-474
THERMAL PROPERTY CHANGES OF ERODED AND REPETITIVELY
LOADED CFC
SCHMALZ, F., D.S. D’HULST, J.G. VAN DER LAAN
NRG, P.O.Box 25, NL 1755 ZG PETTEN
The power handling characteristics of CFC tiles may change after long term exposure to plasmas and repetitive
power fluxes. In ITER such effects may appear in an early operation phase. Samples taken from graphite and
CFC tiles exposed to large plasma fluences in various EU tokamaks are subjected to short pulse high heat
fluxes, chosen below the material ablation limit. Their transient heat load responses are evaluated by applying
Nd:YAG laser pulses of 0.2 up to 10 ms duration, and measuring the surface temperature response with a fast
IR pyrometer. Specimen temperatures are varied between 300 and 1000 K and testing is performed in vacuum.
Power densities are chosen to have peak temperatures in the range of 1200 to 1700 K. In order to quantify the
effect of morphology changes, un-exposed specimens are tested at similar heat loads. The materials shorter
pulse response is more sensitive to near surface properties (< 0.1 mm). In selected cases thermal diffusivity is
measured in addition. ITER CFC grades are subjected to power fluxes that approach marginal erosion
conditions in ELMs. The possible deterioration of the material under repetitive pulses (up to 10^3 pulse at
energy densities up to 1.2 MJ/m²) is investigated. The focus is placed on particular microstructural effects, like
matrix-fibre detachment and enhanced erosion for energy fluxes lower than those causing ‘brittle damage’.
The paper will report on the detailed experimental procedure, thermal property determination and the results of
the specimen damage evaluations.
Corresponding Author:
SCHMALZ, F.
schmalz@nrg-nl.com
NRG, P.O.Box 25, NL 1755 ZG PETTEN
300
- F - Plasma Facing Components.
P4C-F-475
STUDIES OF HEAT CONDUCTION IN LIQUID LITHIUM CAPILLARY
POROUS SYSTEM
AZIZOV ENGLEN, A.G. ALEKSEYEV, V.B. LAZAREV, S.V. MIRNOV V.A. EVTIKHIN, I.E. LYUBLINSKI, A.V.
VERTKOV
1 TRINITI, Troitsk, Moscow reg., 142190 Russia 2 State Enterprise «Red Star» - «Prana-Center» Co, Moscow, Russia
An application of liquid metals to the design of divertor plates is considered now as a promising approach for
the future fusion devices [1-3]. Liquid Lithium (LL) capillary-porous system (CPS) proved to be one of the
most reliable technical solutions, which is able to withstand up to 20 MW/m2 power loading. A number of rail
limiters with LL CPS were developed and tested in the T-11M tokamak starting from 1997 [4-6]. Recent
progress in this R&D program includes a quasi-stationary limiter providing the saturation of temperature
gradient in CPS for the plasma discharges longer than 0.1 sec [6]. One of the most essential problems of CPS
design for the steady state heat loading is providing a reliable thermal contact between liquid and solid parts of
the heat sink, which might be subjected to an occasional destruction under the thermal cycling and heavy local
power loading. The last version of LL CPS limiter installed into the T-11M tokamak has demonstrated an
ability to improve the thermal contact after the long-term uniform thermal heating, and even to recover an
initial situation after the local damage. Some other aspects of heat conduction in the LL CPS under the
powerful loading are discussed also. References 1. L.G. Golubchikov, et al. J. Nucl. Mater., v.233-237 (1996),
667-672. 2. V.A. Evtikhin, I.E. Lublinski, et al., Proc. 16th Int. Conf. on Fusion Energy, Montreal, 7-11 Oct.
1996, Fusion Energy 1997, IAEA, Vienna, 1997, vol. 3, 659-665. 3. V.A. Evtikhin, I.E. Lyublinski, A.V.
Vertkov, et al., Fusion Energy 1998, IAEA, Vienna, 1999, vol. 4, p. 1039-1313. 4. V.B. Lazarev, E.A. Azizov,
A.G. Alekseyev, et al. 26th EPS Conf. on Contr. Fusion Plasma Physics, ECA, 1999, vol. 23I, pp. 845-848. 5.
A.M. Belov, V.B. Lazarev, A.G. Alekseyev, S.V. Mirnov, I.N. Makashin, 28th EPS Conf. on Controlled
Fusion and Plasma Physics, Madeira, Portugal, 2001. 6. V.B. Lazarev, et al, 30th EPS Conference on Contr.
Fusion and Plasma Phys., St.Petersburg, 7-11 July 2003 ECA, Vol. 27A, P-3.162.
Corresponding Author:
AZIZOV ENGLEN
alexag@triniti.ru
TRINITI, Troitsk, Moscow reg., 142190 Russia
301
- G - VESSEL-IN VESSEL ENGINEERING AND REMOTE HANDLING.
P4T-G-4
DESIGN AND DEVELOPMENT TOWARDS A PARALLEL WATER
HYDRAULIC WELD/CUT ROBOT FOR MACHINING PROCESSES IN
ITER VACUUM VESSEL
WU HUAPENG, PEKKA PESSI, HEIKKI HANDROOS, JANNE KOVANEN, (2)LAWRENCE JONES
Department of Mechanical Engineering Lappeenranta University of Technology 53851 Lappeenranta, Finland (2)EFDA
Close Support Unit, IPP-Garching, Boltzmannstrasse 2, D-85748, Germany
Selected also for oral presentation
O4B-G-4
For the assembly of the ITER vacuum vessel sector, precise positioning of welding end-effectors, at some
distance in a confined space from the available supports, will be required, which is not possible using
conventional machines or robots. This paper presents a special robot, able to carry these welding and also
machining processes from inside the ITER vacuum vessel, consisting of a five-degree-of-freedom parallel
robot mounted on a carriage driven by electric motor/gearbox on a rack. The robot carries the machining tool
or welding gun such as a TIG, hybrid laser or e-beam welding gun to weld the inner and outer walls of the
ITER vacuum vessel sectors. The kinematic design of the robot has been optimised for ITER access and a
hydraulically actuated pre-prototype built. The machining force analysis and the optimisation of the machining
processes is discussed in the paper. The welding process with a 3-D seam tracker is introduced. The motion
control of the robot is challenging problem due to the nonlinear behaviour of the mechanical structure and
hydraulic system. A hybrid controller is designed for hydraulics and control system of the robot, including
position, speed and pressure feedback loops to achieve high control accuracy and high dynamic performances.
Finally the experimental results are presented and discussed. Keywords: Parallel robot, ITER vacuum vessel,
Machining/welding, water hydraulic,and Control
Corresponding Author:
WU HUAPENG
huapeng@lut.fi
P.O.Box 20, FIN-53851 Lappeenranta, Finland
302
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-14
BAKING SYSTEM FOR EAST VACUUM VESSEL
Y.L CHENG, Y.T. SONG D.M YAO
Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126,Hefei,Anhui 230031,China
EAST is a medium-size tokamak with super-conducting magnetic field coils. The primary function of the
vacuum vessel are to provide a high quality vacuum for the plasma discharge. Baking technique is applied to
EAST with a view to removing impurities which are mostly due to the release of containants from limiters and
walls during the startup of the plasma discharge. Two baking systems are considered at one time resulting
from the quite different structural characteristics. One is that the vacuum vessel of double-wall structure is
heated by a flow of hot nitrogen gas between the inner and outer shell. The other is an electric heater roiling
around the ports of the vessel. While the systems provide the maximal baking temperature of the vacuum
vessel to be equal to 250ŽC to ensure to achieve acceptable thermal stresses and deformation due to
temperature gradients, we must attach much importance to the non-uniform heating. The baking systems are
presented. Some simulated analyses have been done to make sure to achieve their design specification and to
expect some demolishing factors in the operating process.
Corresponding Author:
Y.L CHENG
songyt@ipp.ac.cn
P.O.Box 1126, Anhui, Hefei, P.R.China, 230031,Institute of Plasma Physics, Chinese Academy of Sciences
303
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-18
LASER MEGAJOULES CRYOGENIC TARGET DEVICES
BRISSET DIDIER, VALERIE LAMAISON (1) GAEL PAQUIGNON (1) ERIC BOULEAU (2) DENIS
CHATAIN (2) JEAN MANZAGOL (2) JEAN PAUL PERIN (2)
(1) Centre Etudes Scientifiques et Techniques Aquitaine /Departement des Lasers de Puissance B.P. 2 33114 Le Barp
FRANCE (2)Commissariat Energie Atomique de Grenoble DRFMC/Service des Basses Températures 17 Rue des Martyrs
38054 GRENOBKLE Cedex 9
LMJ program claims to obtain Deuterium-Tritium (DT) mixture combustion leading to a fusion gain of ten.
The cryogenic targets for inertial confinement, driven by 240 laser beams, are hollow spheres. Their internal
shell are covered with a solid cryogenic fuel layer. The success of DT combustion depends on quality of the
fuel layer geometry. Cryogenic targets must be cooled and kept at temperatures near the triple point (19 K)
with a very good stability (1 mK) for many hours. In the French concept, the targets will be transferred at 22K
+/-2K to the cryotarget positioner using another cryostat wich is carrying target from TargetLab to LMJ
building . Besides that sharp thermal characteristics, the Cryogenic Target Positioner (CTP) displays others
technical challenges like very strong mechanical specifications. The CTP deals with the target handling and
positioning in the center of the 5 m radius experimental target chamber with a precision of few microns. The
target transfer and positioning must be entierely remote controlled and done under vacuum. In order to validate
our current CTP conception, we have manufactured a One Scale Cryostat Demonstrator to confirm all CTP
thermal challenges, such as sharp thermal regulation, cooling autonomy and cryogenic target transfer. First
results obtained with this prototype will be presented.
Corresponding Author:
BRISSET DIDIER
didier.brisset@cea.fr
CEA/CESTA BP2 33114 LE BARP FRANCE
304
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-29
IRRADIATION TESTS ON WATER HYDRAULIC COMPONENTS
T. HERNANDEZ, AND E.R. HODGSON
Euratom / CIEMAT Fusion Association, 28040 Madrid, Spain
Remote handling operations in ITER will require the use of hydraulic systems for lifting and moving activities,
such as those envisaged for the CMM (Cassette Multifunctional Mover) and the CTM (Cassette Toroidal
Mover). These hydraulic systems will make use of high pressure water up to 210 bar, rather than oil. Hence the
polymer materials employed as seals, glide rings, and wipers on the pistons and cylinders will be subjected to
gamma irradation in the presence of water or high humidity. It is therefore necessary to study the combined
effect of radiation and water on the degradation of the polymers, as well as possible enhanced water corrosion
due to gases such as chlorine or fluorine released from the irradiated polymers. The use and trade names of the
materials which have been tested are: Piston seal: Dehoplast PE-1000 (UHMW-PE). Glide ring: Merkel Hard
fabric HGW HG517 (Freudenberg Simrit). Wiper: Merkel Novathan 95 AU V149 (polyurethane). Gamma
irradiations have been performed in the CIEMAT 60Co pool facility (Nayade), which allows irradiation in a
closed sample chamber at a controlled temperature and in a controlled atmosphere. Tests were carried out on
the materials to 1.0, 3.3, and 10.0 MGy at 9 Gy/s, 300 K to evaluate the modification / degradation of the
physical properties of the polymers. The irradiations were performed in two different environments: dry
nitrogen and in deionised water. In addition control samples were put in deionised water without irradiation
during the time needed to reach 1, 3.3, and 10 MGy in order to compare the effects of water alone, and water
plus radiation. Following the tests microstructure, microhardness, chemical analysis, and water absorption
were examined. Results for UHMW-PE (Ultra high molecular weight- Polyethylene) seals, NBR
(Acrinolnitrile-butadiene rubber) O-rings and Polyurethane wiper rings demonstrated no notable degradation
up to 10 MGy. However, by 10MGy the hard fabric material for the glide ring (piston bearing) exhibited
undesirable levels hardening, production of fluorides in the water and the fibre matrix appears to have been
dissolved.
Corresponding Author:
T. HERNANDEZ
teresa.hernandez@ciemat.es
Euratom / CIEMAT Fusion Association, 28040 Madrid, Spain
305
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P4T-G-53
EXPERIMENTAL RESULT OF THE LASER IN VESSEL VIEWING AND
RANGING SYSTEM (IVVS) FOR ITER
NERI CARLO, L. BARTOLINI, A. COLETTI, M. FERRI DE COLLIBUS, G. FORNETTI, F. POLLASTRONE, M.
RIVA, L. SEMERARO
Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome,Italy
A prototype of Laser in Vessel Viewing and Ranging System (IVVS) was developed at ENEA laboratories. It
is based on an amplitude modulated (AM) laser radar specifically designed to withstand with the severe ITER
conditions. The system is able to perform, at the same time, viewing&ranging of in-vessel surface. The target
is scanned using a radiation resistant laser beam deflection system coping the ITER thermonuclear and
mechanical constraints. The most critical parts of the system, which is the fiber optic optical encoder, has been
successfully tested by SCK-CEN laboratories in Mol under radiation at 15 KGy/h up to a dose of 2.47 MGy. A
series of viewing and ranging tests have been performed on the system to better evaluate its main
characteristics that can be resumed in auto-illumination, large field of view, zoom capability, range
measurement capability, high dynamic range (thousand of grey levels can be resolved), relative immunity on
speed fluctuation of the scanning mechanism. It was verified the large field of view of the system that is able
to take images in very wide angles (quasi-spherical images) as well as in restricted areas. Resolution charts
have been scanned to evaluate the maximum resolution of the system, which has been able to distinguish rows
of 0.55 mm at a medium distance of 3.5 m; the results was in accordance with the theoretical one <1mm in the
range 2-7 m although speed fluctuation of about 20% has been observed. Test has been performed on the
ranging capability of the system using a testing sample made of three group of stairs of 10 mm, 5 mm and 1
mm of step; the obtained image allowed to see well all the stairs groups both in amplitude and in range. A
Standard Deviation of the range measure of 320 µm with a 3 ms stay time of the laser beam was reached
comparing the measurement with the CAD representation of the sample. The paper shortly describe the overall
architecture of the system then the radiation test performed on the more critical components and in more detail
presents the experimental viewing and ranging results obtained, showing the main characteristics and the
advantages of the system.
Corresponding Author:
NERI CARLO
neri@frascati.enea.it
Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome,Italy
306
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P4T-G-56
ITER VACUUM VESSEL SECTOR MANUFACTURING
DEVELOPMENT IN EUROPE
LAWRENCE P JONES, ALDO BIANCHI (1), ALAIN CROS (2), ENRICO DI PIETRO (3), BENOIT GIRAUD (2),
KIMIHIMO IOKI (4), LUBOMIR JUNEK (5), BRUNO PARODI (1), MICHAEL PICK (3), GIAN-PAOLO
SANGUINETTI (1), RICHARD TIVEY (4), YURI UTIN (4)
(1) Ansaldo Ricerche - Corso Perrone 25, 16161, Genova, Italy, (2) Framatome ANP - 10 rue Juliette Récamier, 69456,
Lyon, France, (3) EFDA CSU - Garching, (4) ITER JCT - Garching, (5) Inst. of App. Mechanics Brno - Veveri 95, 61139,
Brno, Czech Republic
The ITER Vacuum Vessel provides the first Tritium and Vacuum boundary and supports the first wall blanket
and divertor modules, the attachment requirements of which complicate the construction of the vessel and
place manufacturing tolerances on the VV several times smaller than usual in relation to it’s large size so that
there is a risk of rejections after manufacture due to out-of tolerance for the 9 Vacuum Vessel Sectors.
Utilising large bracing fixtures stiffer and heavier than the vessel itself, European Industry has proposed a
manufacturing route for the Sector construction, and the ITER International Team has accepted this method as
the reference. However, the achievement of the required tolerances remains challenging and has to be
validated prior to their procurement. The central part of this validation is the placement of a contract for the
procurement of a full-size, poloidal segment, consisting of a 40 degree, 5 metre high, 20 Ton part of the
inboard section, fabricated according to the manufacturing route, including bracing fixtures, welding
applications, restraint effects, and fit-up aspects. A steel beam structure with stiffness comparable to the
missing part of the sector is included, as this has an important influence on the distortion, which will be
measured at the end of the fabrication. The finished segment is used as the basis for a mock-up that simulates
the joining of two sectors. Since the goal of this program is to be able to use the model to extrapolate the
segment distortions to the actual ITER VV sector, numerical simulations by a European Association, using the
SYSWELD program, is used with a new module, incorporating local models of instrumented welding
coupons, the results of which are included in a series of simplified global models. This paper describes the
results of the manufacturing development programme so far.
Corresponding Author:
LAWRENCE P JONES
lawrence.jones@tech.efda.org
EFDA CSU - Boltzmannstr. 2, 85748, Garching, Germany
307
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P4T-G-71
STRUCTURAL UPGRADE OF IN-VESSEL CONTROL COIL ON DIII D*
ANDERSON, P.M., A.G. KELLMAN, E.E. REIS
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
For most of 2003, DIII-D operated 12 new in-vessel outer wall mounted control coils. The single turn,
rectangular coils are mounted in 2 levels of 6 coils each. The coils were used for many experiments such as
suppression of the resistive wall mode, for correction of magnetic field imperfections and for creation of an
ergodic edge magnetic field for the suppression of edge localized modes. During operation with a maximum
current of 4.5 kA at 100 Hz, one coil developed a leak through the stainless barrier that separates the nitrogen
blanketed insulated conductor from vessel vacuum. A pair of coils was taken out of operation for the last
month of the year. This paper describes the failure investigation, design, analysis, component testing, repair,
system testing and new interlocks for the system that will see significant use in 2004. The crack in the stainless
barrier was attributed to low cyclic fatigue related to operation of the coil at 100 Hz, a frequency near the
vertical natural frequency of the coaxial lead. Finite element analysis (FEA) after the failure showed that
electromagnetic forces on the single conductor section were sufficient to excite the coaxial lead in a vertical
mode. The crack likely developed in less than a second. Repair options were limited. The coils are mounted to
the walls and removal of PF tiles was discouraged in order to minimize the repair time. Welding on the coil
was limited in order to protect the internal polyamide insulator from overheating. Repairs included: 1) seal the
leak in the faulted coil, 2) increase the stiffness of the single conductors near the coaxial transition and 3)
significantly increase the first natural frequency of the coaxial leads to allow operation to 1000 Hz. In-vessel
vibration testing was done at each stage of repair to compare the natural frequency of the three types of leads
with that determined by FEA models. Verification testing was done prior to vessel closure. The test included
temporary installation of field tolerant strain gages to monitor strain in the stainless for comparison with model
results. Permanent vacuum vessel port deflection monitors were added with the hope that excessive lead
vibrations could be detected by port deflections for interlock purposes. All 12 coils were successfully repaired,
upgraded and test results are encouraging. *Work was supported by the U.S. Department of Energy under DEFC02-04ER54698.
Corresponding Author:
ANDERSON, P.M.
anderson@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
308
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-89
MANUFACTURING OF CRYOSTAT FOR EAST SUPERCONDUCTING
TOKAMAK
YU JIE,
EAST superconducting Tokamak is under fabrication and in procurement phase at Institute of Plasma Physics,
Chinese Academy of Sciences. One of the main parts of ESAT is the cryostat which provides the thermal
protection of the coil system. The cryostat consists of a cylindrical section bolted to dished lid wall at top and
to base plate at bottom by flanges with special C clamps. The lid wall of the cryostat is a dished configuration
for reasonable stress distribution. The support of the cryostat on the base has been designed for transferring the
loads of the vacuum vessel and magnets to the basis of the machine test hall directly. All parts of cryostat
withstand the design basic loads, which include external pressure at most operating condition (1bar), dead
weight, electromagnetic forces and seismic load. The manufacturing of the cryostat is under way and will be
finished in July, 2004. Totally sixteen horizontal ports and thirty-two vertical ports are designed for the
requirement of diagnostics and operation. With physical diagnostics and test specifications, three types of
horizontal ports with different shapes and size are needed, in which three of them are capable for tangential
neutron beam injection and physical diagnostics. In addition, there are eight horizontal man ports for future
maintenance. There are 16 vertical access ports and four maintenance ports (the same four ports in the down
section of middle ring) in lid wall. There are sixteen horizontal ports to the machine vacuum vessel at the
machine equator. There are two types sixteen vertical ports, eight cryogenic ports and one helium exhaust port
on the base plate. The cryostat is made of 304L stainless steel. The cryostat is 7592mm outside diameter and
7095mm height (not including main support). The cryostat is manufactured by Shanghai Boiler Works, Ltd
(SBWL). The inside wall radius is 3661mm and the height of cylindrical section in middle ring 4460mm. The
lid is made of 28mm 304L stainless steel, which minimum thickness is 25mm. The non-standard dished head
spherical and knuckle radii are 10000mm and 850mm, respectively. The dished lid was fabricated at the end of
2002 by using a set of head ramming machine and a set of head flanging machine (BOLDRINI) imported from
Italy with working diameter from 1800mm to 8000mm.
Corresponding Author:
YU JIE
yujie@ipp.ac.cn
P.O.Box 1126, Hefei Anhui 230031, P.R.China
309
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P4T-G-117
SPECIAL BLANKET DESIGN IN THE NB REGION OF ITER
ELIO FILIPPO, K. IOKI , Y. UTIN, M. MORIMOTO
ITER International Team Boltzmannstrasse 2 85748 Garching Deutschland
The ITER blanket is a 45 cm thick nuclear shield mounted on the vessel as 440 modules with regular shape
and a weight below 4.5 tonnes. The modules have a straight profile 85-120 cm long allowing the production of
the first wall (FW) in flat panels, 25-40 cm wide, four per module. To maintain a trapezoidal front view, the
boundaries between the modules have been aligned with the edges of the equatorial and upper ports of the
vessel. The design of the FW panels, the main stainless steel body and the mechanical attachment to the vessel
wall has been developed and tested for this configuration. Unfortunately the 3 Neutral Beam (NB) openings
violate the 20 cyclic symmetry of the regular ports and would require a different blanket segmentation. These
openings are smaller and cross the FW in the plane of the toroidal field coil. In the past the NB module design
lacked a satisfactory attachment on the vessel and was not fully integrated with the standard blanket portions.
The requirements of the NB openings have been reviewed and compared with the design constraints of the
vessel, of the blanket cooling manifolds, of the module and its attachment. A new layout has been developed
which appears to be a good compromise for all components and a good basis for the detailed design. The
height of the NB opening has been squeezed symmetrically from 136 to 116 cm resulting in a higher, though
still acceptable beam heat flux on the sides of the duct. The equatorial portion of the blanket cooling manifolds
has been curved toroidally around the beam. The blanket module on the side of the NB opening exposed to the
plasma has been extended 80 cm along the inside of the port. Thus it has a wider region for the attachment and
completely protects the vessel corner from the nuclear radiation. The poloidal segmentation is now 2 modules
as it is elsewhere between the equatorial ports, the toroidal spacing is also close to the regular angular pitch of
10 with a deviation imposed by the slanted sides parallel to the NB axis. All these benefits require a new
effort to develop the corner blanket module, which is special and exposed to nuclear radiation and surface
heating on two sides. The paper explains the blanket design evolution in the NB region, lists the requirements
of all interacting components, presents their individual design and discuss it. The implications for the handling
and the feasibilty of a separable FW with its reduced waste are also reported.
Corresponding Author:
ELIO FILIPPO
eliof@itereu.de
ITER Team, Boltzmannstrasse 2, 85748 Garching, Deutschland
310
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P4T-G-137
USE OF ELECTRONIC AND OPTOELECTRONIC INDUSTRIAL
SYSTEMS FOR MAINTENANCE TOOLS OF ITER FUSION
EXPERIMENTAL REACTOR
GIRAUD ALAIN, MARCO VAN UFFALEN(1) FRANCIS BERGHMANS(1)
(1) SCK CEN, Department Instrumentation Boeretang 200, B-2400 Mol, Belgium
The environmental constraints encountered by the maintenance tools of the future ITER Fusion Reactor such
as CMM, SCEE, DTP and DRP, could reach a few MGy of total dose irradiation while temperature could rise
up to 150 C. The necessary remote handling systems will generate a huge number of wires to connect the
sensors and actuators to the control room. The use of embedded electronic and optoelectronic systems could be
an appropriate solution, for the future designers, to reduce the size of umbilical and connectors, facilitate their
movements and limit inside failures. Involved since fifteen years in the developments of industrial electronic
systems for civilian nuclear activities (AREVA, EDF, …), our laboratory has gained its knowledge in the
permanent understanding of the radiation behavior of components and the optimized design of electronic
architecture to extend the lifetime of on board systems. In the same way, SCK laboratory has developed an
similar approach for optoelectronic components and optical fibers. Taking into account the common use of
sensors with analog output signals, a experimental data link was built in 2002 for communications between an
embedded sensor and the control room. A mock-up including electronic and optoelectronic components was
designed to digitalize the analog output signal and transfer it, after an electric to optical conversion, through
optical fiber. The expected tolerance to the environment was established without significant degradation of
converted signals. The continuous survey of emerging “off the shelf” technology allowed us to propose now a
pre-prototype of a data multiplexer able to generate a 16 bits frame. Data inserted represent the digitalization
of analog output coming from well-known sensors like LVDT or resolvers, limit switches or proximity
sensors. The transfer to the control room can be done either by a bifilar or optical support at a frame rate of
32kHz which meets most of the remote processes needs. Validation to radiative environment was performed
without any failure. The regularity of all analog conversions was correctly assumed. The final prototype could
be proposed to demonstration and final validation by the end of this year. Machine designers and end-users
will be provided timely with appreciable tools to limit wires and simplify umbilical problems by proceeding,
all along ITER life, to a regular exchange of “electronic black boxes”.
Corresponding Author:
GIRAUD ALAIN
alain.giraud@cea.fr
CEA/DRT/LIST/DTSI/SARC Bat 451 CEA/Saclay 91191 Gif-sur-Yvette FRANCE
311
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-199
NON-DESTRUCTIVE TESTING OF BONDED STRUCTURES FOR
PLASMA FACING COMPONENTS
ONOZUKA, MASANORI, K. KIKUCHI, A. KIRIHIGASHI, Y. ODA, AND K. SHIMIZU
Mitsubishi Heavy Industries, Ltd. Kobe Shipyard & Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe, 652-8585
Japan
Because of material characteristics requirements and adequate heat removal capability, bonded structures have
been employed for plasma-facing components in the International Thermonuclear Experimental Reactor. To
ensure the structural integrity of the bonded structures, non-destructive testing is to be conducted. Amongst the
various types of non-destructive testing, ultra-sound testing (UT) was examined, with emphasis on UT
inspection of the first-wall panel of the blanket module. Three test samples simulating the first-wall panel were
fabricated. Two plates made of Cu-Cr-Zr alloy were bonded with stainless steel (SS) cooling pipes that were
inserted between the two plates to form a heat sink structure. The heat sink structure was then bonded to a SS
structural block. Thus, there are three bonded interfaces: the first between the Cu alloy plates, the second
between the Cu alloy plate and the SS block, and the third between the Cu alloy plates and the SS pipe. In the
test samples, several artificial defects were applied along the bonding interfaces. Three types of UT probes
have been tested. A vertical UT probe and a phased array UT probe were used to detect defects between the Cu
alloy plates, and between the Cu alloy plate and the SS block. Both the probes were applied on the Cu alloy
surface or on the SS block surface. To detect defects along the SS pipes, a beam-focused type UT probe has
been applied. The focused-type probe was inserted into the pipe for detection. In addition, to attain a better
signal to noise ratio (S/N), a noise reduction technique has been applied using digital data processing. Using
the above UT probes, artificial defects as small as 2 mm in size have been successfully detected at a S/N ratio
of more than 2. Details of the study will be presented at the symposium.
Corresponding Author:
ONOZUKA, MASANORI
masanori_onozuka@mhi.co.jp
Mitsubishi Heavy Industries, Ltd. Nuclear Systems Engineering Department, Minatomirai 3-3-1, Nishi-ku,
Yokohama, 220-8401 Japan
312
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-240
VERTICAL DIPLACEMENT EVENTS SIMULATIONS FOR TOKAMAK
PLASMAS
PACCAGNELLA ROBERTO, T.BOLZONELLA, M.CAVINATO, S.ORTOLANI, G.PAUTASSO(1), W. SCHNEIDER
(1), V.LUKASH(2), H. STRAUSS(3)
(1) Max Planck Institute, IPP (Garching, Germany) (2)RRC Kurchatov Institute (Moscow, Russia) (3)Courant Institute for
Mathematical Sciences (NY, USA)
In this paper we study the so called Vertical Displacement Events (VDEs) in an elongated and diverted
tokamak plasma. The study is carried out using two numerical codes: DINA [1] which is a nonlinear magnetohydro-dynamic (MHD) 2D code (assuming plasma axi-symmetry) evolving the plasma through equilibrium
states which satisfies the toroidal Grad-Shafranov equation and which takes into account of the
electromagnetic interaction with metal walls and external coils; M3D [2] a 3D multi-level toroidal code which
evolves in time the full MHD equations and takes into account a resistive wall surrounding the plasma by
matching the internal plasma magnetic field with the external vacuum solution. In this paper we compare the
time evolution of a VDE for the axi-symmetric (DINA) and the non axi-symmetric (M3D) cases starting from
the same initial equilibrium. We analyze ITER-like equilibria and also some equilibria relevant for the
ASDEX-U experiment. For the simulations in the M3D code a “virtual casing” method is used with the coil’s
currents held fixed at their initial values, while in the DINA code the coils currents can or cannot be constant
in time during the evolution. In the case of ASDEX-U a comparison of the experimental data with the 2D
simulation results is performed while for the 3D case the study is focusing on the identification of suitable
conditions able to represent the effect of the passive structures. This study is particularly important in order to
estimate the symmetric (DINA) and non axi-symmetric (M3D) structure of halo currents during VDEs. This
can have an import impact on the mechanical design of future experiments like ITER. The requirements for the
VDE control system in such devices can be also affected by the computations presented here. [1 ] R.R.
Khayrutdinov, V.E. Lukash, Studies of Plasma Equilibrium and Transport in a Tokamak Fusion Device with
the Inverse-Variable Technique, Journal of Comp. Physics 2, 106, (1993) [2 ]Park W., et al., Phys.Plasmas 6,
1796 (1999)
Corresponding Author:
PACCAGNELLA ROBERTO
roberto.paccagnella@igi.cnr.it
Consorzio RFX Corso Stati Uniti 4 35127 Padova
313
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P4T-G-249
DESIGN OF THE ITER HOT CELL BUILDING
J HAYWARD (1), D MAISONNIER (1) O ASUAR (2) T FISHER (3) T EURAJOKI (4) AND JÖELLE ELBEZ-UZAN
(5)
(1) EFDA Close Support Unit, Garching; (2) EFET, IBERTEF-EA; (3) EFET, NNC; (4) EFET-Fortum; (5) Euratom-CEA
Association
The Hot Cell building is a reinforced concrete building, which provides the facilities required for a variety of
operations on activated in-vessel components and systems. Due to the erosion of the plasma-facing
components by the high thermal loads, the divertor will need to be replaced and upgraded several times during
the life of ITER. To minimise the amount of activated waste, the divertor is of modular design and based on
the use of reusable cassettes, which can be removed from the Tokamak to be repaired and refurbished with
new plasma-facing components. The in-line repair and refurbishment of divertor cassettes is the main
operational requirement of the Hot Cell, but the repair and refurbishment of other Tokamak components,
including diagnostic plugs, blanket modules and RF heating port plugs, are also part of the specified Hot Cell
operations. Additionally, the Hot Cell has to process and store waste accumulated from the Tokamak during its
operational lifetime. At present, the size and layout of the building is determined principally by the
maintenance requirements and to provide all the facilities necessary for equipment storage, repair and testing,
exchange and maintenance of remote handling tools, and waste processing and storage. A number of studies
and reviews of remote maintenance activities have resulted in a simplification of the remote handling
methodologies and of the concepts for the refurbishment tasks. The design of the Hot Cell building must also
be reviewed with respect to the applicable regulatory codes, the possible requirement for a later extension of
the facility to cater for the dismantling of the in-vessel component systems during the de-activation phase of
the project, and other specific functions, such as in-vessel component docking, dust cleaning, transfer cask
storage, atmosphere confinement control, and atmosphere de-tritiation. The paper will describe the results of
these reviews and proposed changes to the design to meet both the functional and regulatory requirements.
WITHDRAWN
Corresponding Author:
J HAYWARD (1)
jim.hayward@tech.efda.org
EFDA Close Support Unit, Max-Planck-Institut für Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching-beiMünchen, Germany
314
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-264
RECENT DEVELOPMENTS TOWARDS ITER 2001 DIVERTOR
MAINTENANCE
PALMER JAMES, MIKKO SIUKO(1) PIETRO AGOSTINI (2) ROLAND GOTTFRIED (3) MICHAEL IRVING (2)
ALESSANDRO TESINI (4) MARCO VAN UFFELEN (5)
((1) Tampere University of Technology, Tampere, Finland (2) ENEA CR Brasimone, Bologna, Italy (3) Framatome ANP
GmbH, Erlangen, Germany (4) ITER International Team, Naka, Japan (5) SCK-CEN, Mol, Belgium
The divertor assembly for ITER consists of 54 rail-mounted cassettes located in the bottom region of the
vacuum vessel. Due to the erosion of the plasma-facing components and the possible need for improving the
divertor design, its periodic replacement is foreseen a number of times during ITER’s 20 year operational
lifetime. In moving from the ITER’98 to the more compact ITER 2001 design, although the general principles
of divertor RH remained intact, the detailed design of almost all the divertor handling equipment had to be
significantly changed, mainly due to the reduction in space between the cassettes and the inner wall of the
vacuum vessel. This feature prevents the use of the simple “trolley-like” cassette carriers developed for
ITER’98 (and modelled in the Divertor Test Platform (DTP) at Brasimone), but necessitates the use of
cantilevered cassette handling using a more complex device known as the “Cassette Multifunctional Mover”
(CMM). In this new approach the vertical position of the cassette during its passage along the vacuum vessel
duct is no longer simply related to a fixed set of radial rails but has to be continually adjusted in free space by a
serial chain of robotic joints. Added to this, the nominal clearances between the cassette and vacuum vessel
duct are only 30 mm which sets extreme demands for the mover control system. Since early 2003 the EU
Participant Team has been engaged in the detailed design of the CMM together with its set of specialised endeffectors. The efficiency of this process and integration of the CMM design into that of the ITER machine,
have been greatly enhanced by the use of state-of-the-art virtual reality and virtual prototyping techniques
using Igrip and ADAMS software. During the same period a new RH mock-up facility, designated DTP2, has
been designed and specified. Its main purpose will be to allow demonstration and refinement of the CMM
design and related operational procedures in preparation for procurement of the actual RH equipment to be
used in ITER. This paper will briefly describe the current ITER divertor replacement rationale, report on the
latest cassette mover designs and outline the nature and objectives of the new DTP2 test facility.
Corresponding Author:
PALMER JAMES
jim.palmer@tech.efda.org
EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany
315
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-293
MANAGEMENT OF A WATER LEAK ON ACTIVELY COOLED
FUSION DEVICES
SAMAILLE FRANK, M. CHANTANT, D. VAN HOUTTE, J.J. CORDIER, L. GARGIULO
Association Euratom-CEA, CEA/DSM /DRFC, CEA- Cadarache, 13108 Saint Paul Lez Durance, France,
ITER will be the most important machine equipped with actively cooled Plasma Facing Components (PFCs).
In case of abnormal events during a discharge, the PFC will be submitted to localized transient phenomena
(high power densities, run away electrons, etc ), leading, in the worst case, to the degradation of the PFC wall
and possibly to a water leak. In any case, a leak will have important consequences for the PFCs and
equipments located in the vacuum vessel or connected to the ports such as seals, pumping systems or
diagnostics. A great experience of these events has been gained at Tore supra since more than 10 years and it
will be useful for the next step machines. During 2002 and 2003 experimental campaigns, several leaks
occurred at Tore Supra. In the most important one, 2000 liters of water were spilled in the vacuum. During this
major event, specifics actions have been done to limit the damages and especially to preserve the aluminum
seals. In the first seconds after a leak occurs, if possible, a fast and reliable research of the leaking circuit has
to be carried out. The isolation valves of the circuit are then closed in order to reduce the inside pressure in the
lines and to limit the water vapor flow into the vacuum vessel. In the case of a large leak, all the circuits of the
cooling loop connected to the Tokamak which are located in the Torus Hall are drained off by using a wired
safety device. A drain off procedure has been defined and it is continuously improved. At the present time, the
drain of the PFCs fed by the upper part circuit is not fully satisfactory because of several lines connected in
parallel. Preliminary experiments have been performed to improve it and led to encouraging results. Once the
leaky PFC has been emptied, some effort is required to evacuate the water from the vacuum vessel. After
removal of water by gravity, and then, by baking and pumping in the vessel, the identification and repair of the
leaky PFC start. The quick restart of the systems after the most severe leak of September 2002 with a pumpout without any air leak at the level of one hundred Aluminum seals confirms the convenience of the
performed actions. The paper will present the description of the procedures applied to put the system in safety
depending on the gravity of the leak. It will also present the methods used at Tore Supra to drain-off the
primary loop circuits and to determine the leaky PFC.
Corresponding Author:
SAMAILLE FRANK
frank.samaille@cea.fr
Association Euratom-CEA, DSM / DRFC, CEA- Cadarache, 13108 Saint Paul Lez Durance
316
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P4T-G-296
DESIGN PROGRESS OF THE ITER VACUUM VESSEL AND PORTS
UTIN YURI, V. CHUYANOV(1), F. ELIO(1), K. IOKI(1), L. JONES(2), V. KOMAROV(3), E. KUZMIN(3), M.
MORIMOTO(1), M. NAKAHIRA(4), G. SANNAZZARO(1)
(1) ITER IT, Boltzmannstr. 2, 85748 Garching, Germany (2) EFDA, Boltzmannstr. 2, 85748 Garching, Germany (3) NTC
"Sintez", Efremov Inst., 189631 St. Petersburg, Russia (4) JAERI, Naka Fusion Research Establishment, Naka, Ibaraki, 3110193, Japan
The ITER vacuum vessel (VV) is a torus-shaped double-wall structure with stiffening poloidal/toroidal ribs
between the shells. The VV main function is to provide the high-vacuum and primary safety confinement
boundary. The vessel also supports the blanket and the divertor components. Along with the in-vessel
components, the VV provides radiation shielding – the neutron heat is removed by the water circulating
between the shells. To provide access inside the vessel for auxiliary plasma heating, diagnostics, vacuum
pumping and other needs, the VV is equipped with upper, equatorial, and lower ports. Approaching the ITER
construction phase, the VV design has been improved and developed in more detail with the focus on
simplified manufacture and reduced cost. Options of the general fabrication scheme have been considered in
cooperation with the industrial companies and the design has been updated in conformity with the main
manufacture requirements and recommendations. To simplify the design, the inboard triangular supports of the
blanket modules have been eliminated and the design of the outboard supports refined. Another important
improvement is that the VV supporting system has been modified to provide better access to the main
supporting components after assembly of the machine. For some ports, a single-wall construction will be used
at a certain distance from the main vessel, where the neutron load is less intense. This approach simplifies the
port manufacture and maintenance. For the upper and equatorial ports, the in-port space is occupied by an
integrated subassembly - the port plug, which, apart from its functional purposes, provides the
vacuum/pressure boundary for the in-vessel volume. Special attention was paid to the design of the supporting
and sealing components between the plug and the port with the focus on improved structural performance and
maintenance. For further cost reduction, the number of large lower ports is halved, with the ports between
every other toroidal field coil. Where there are no ports, there are only pipe feedthroughs and local small
penetrations. The VV must withstand those loads directly induced in the vessel and those transmitted from the
in-vessel components. Based on the performed structural analyses, additional reinforcements have been
incorporated into the main vessel/ports where required. Details of the current VV design and results of the
related analyses are reported in this paper.
Corresponding Author:
UTIN YURI
utiny@itereu.de
ITER IT, Garching JWS, Boltzmannstr. 2, 85748 Garching, Germany
317
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-348
1200 MM BORE VOLTAGE BREAK OF THE NB DUCT FOR KSTAR
B.J. YOON, T. NAGAYAMA(1) S.R. IN(2) B.H. OH(2)
(1)Hitachi Haramachi Electronics, Hitachi-shi, 317-0072, Japan (2)Korea Atomic Energy Research Institute, Daejeon, 305353, Korea
The beam duct connecting the NBI system to the KSTAR (Korea Superconducting Tokamak Advanced
Research) vacuum vessel consists of a large gate valve, a voltage break, transition tubes, and beam stoppers.
The voltage break keeps peripheral devices safe from the potential difference of higher than 10 kV generated
between the NBI and the torus during plasma disruptions and power faults in superconducting magnets and
NBI high voltage system. The voltage break is composed of a ceramic ring, a bellows and fitting flanges. The
voltage break is mechanically very delicate component because it must accommodate thermal and mechanical
relative displacements of the vacuum vessel side and NBI side structures, occurred due to the system baking
and misalignments in the assembly. Therefore, the bellows should be flexible enough to absorb a shift of a few
cm in the axial direction and ~5 mm in transverse and be rigid to withstand the atmospheric pressure exerted
on sidewalls of convolutions. The bellows is designed to be a welded type and have a size of 1200 I.D, 75 mm
width, 14 mm pitch and 15 convolutions to fulfil above requirements. The ceramic ring made of alumina
(Al2O3) is the key part of the voltage break which has a design breakdown voltage of 30 kV. The ceramic ring
has dimensions of 1200 mm I.D., 50 mm thickness and 45 mm width. The ceramic ring is brazed with
KOVAR sleeves on both flat sides, and the sleeves are welded to the bellows and the flange assembly. The
ceramic ring is the most fragile part in the voltage break at the bonding boundary between the sleeve and the
alumina ring. There has been no successful experience of fabricating a 1200 mm bore alumina ring so far in
the world. At the first step a ceramic ring is formed with the alumina powder by pressing and sintering at the
Kyocera Kagoshima factory, and then bonded to KOVAR sleeves using the active metal brazing method in the
vacuum brazing furnace at the Mitzubishi Hiroshima factory. The welding of bellows and flanges to the
ceramic ring is carried out at the Hitachi factory. The entire procedure is managed by Hitachi. The surface of
the ceramic ring is checked with the ink penetration method to find cracks before the brazing procedure. After
brazing the ceramic ring is checked with the eye and leak-tested. The voltage break assembly is tested by
pressurizing inner volume of the break up to 1.5 atm. The strength between the ceramic ring and the KOVAR
sleeve is expected to be more than 80 MPa.
Corresponding Author:
B.J. YOON
bjyoon@kaeri.re.kr
Korea Atomic Energy Research Institute, Daejeon, 305-353, Korea
318
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-355
MANUFACTURE OF THE PLASMA VESSEL AND THE PORTS FOR
WENDELSTEIN 7-X
REICH, JENS, WILLI GARDEBRECHT (1) BERND HEIN (1) BERND MISSAL (1) JOERG TRETTER (1) FRANZ
LEHER (2) STEFANO LANGONE (3)
(1) Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491
Greifswald (2) MAN DWE GmbH Deggendorf, Werftstraße 17, D-94469 Deggendorf (3) Romabau-Gerinox AG,
Fohlenweide, CH-8570 Weinfelden
Selectesd also for Oral Presentation
O4B-G-355
WENDELSTEIN 7-X (W7-X) is a superconducting helical advanced stellarator which is presently under
construction at the Greifswald branch of IPP. Thermal insulation of the 70 coils requires a cryostat. It is being
composed of a plasma vessel, an outer vessel, ports to observe and heat the plasma, cooled shields and
multilayer insulation. The German company, MAN-DWE, is responsible for manufacture of the plasma vessel,
the outer vessel and the thermal protection. The Romabau-Gerinox AG delivers the ports. Following the
symmetry of the magnetic configuration the cryostat is composed of five almost equal modules. The shape of
the plasma vessel has to closely follow the twisted shape of the plasma and has a cross section which
continuously varies between triangular and bean shape. The plasma vessel is composed of 10 half-modules.
Each half-module is again divided into two sectors to allow stringing of the coils during assembly. For each
half-module 20 steel rings are precisely bent to the required shape and carefully welded to represent the
changing cross-section of the plasma vessel. For local areas, which cannot be approximated by bending steel,
sheets are fitted by hot forming. By end of March the four sectors required for the first module of the plasma
vessel were delivered. The contours of the sectors were measured by laser tracker system and met well the
given narrow tolerances. Vacuum tightness of the welds was checked by an integral helium leak test of each
whole sector. Precise cutting of the holes for the ports was performed by water jet technique. Water pipes
around the outside of the vessel allow its temperature to be controlled during plasma operation and for bake
out. A total of 299 ports are used for evacuating the plasma vessel for plasma diagnostics and plasma heating
and for feeding supply lines and sensor cables. Cross sections of the ports range from 100 mm circular to 1000
x 400 mm rectangular. Bellows balancing movements of the plasma vessel during bake-out and final
adjustment. All ports are surrounded by water pipes to control their temperature. By spring of this year 60
ports have been delivered. The paper will summarise the design activities and give a short description of the
fabrication status of the main cryostat components.
Corresponding Author:
REICH, JENS
jens.reich@ipp.mpg.de
Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstraße 1,
D-17491 Greifswald, Germany
319
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-360
DYNAMIC IDENTIFICATION OF THE HYDRAULIC ITER MAESTRO
MANIPULATOR - RELEVANCE FOR MONITORING
BIDARD CATHERINE, C. LIBERSA (1) D. ARHUR (1) Y. MEASSON (1) J.-P. FRICONNEAU (1) J.-D.
PALMER (2)
(1) Robotics and Interactive Systems Unit - CEA, BP 6, F92265 Fontenay-aux-Roses Cedex, France (2) EFDA CSU
Graching, Boltzmannstrasse 2, 85748, Graching, Germany
Maintenance tasks at ITER divertor level requires use of powerful Force feed back Remote Handling device
such as hydraulic Manipulator. CEA in collaboration with CYBERNETIX and IFREMER has developed the
hydraulic manipulator MAESTRO (Modular Arm and Efficient System for TeleRObotics). Force control of
the robot allows for force-reflective telemanipulation, which is required for better manipulation and sensing of
the environment during remote operation tasks. When considering ITER vacuum Vessel conditions, radiation
level will reduce significantly the possibility to operate remote tools with good vision feed back. Therefore,
model based monitoring of the manipulator will be required during operational period to enhance feedback to
the operator. Therefore, it is required to control that the manipulator inside the Vacuum Vessel is well
operating and to warn the operator of possible failures. Without adding any sensors, it is possible to monitor
the relation between the robot torques and trajectory. In this paper we present the experimental identification
of the parameters of the dynamical model of the MAESTRO arm. The control of joints was done on the
MAESTRO arm with flow control servo-valves. The axes torques were derived from pressure measurements
in joints chamber, and the axes positions measured by resolver sensors. The first part presents the experimental
set-up and trajectories. The second part deals with the joints friction models: nonlinear viscous behaviour was
observed and explained by possible non laminar effects of flow inside the hydraulic joints. One axis with
multiple seals showed Stribeck effect. The third part presents the identification of the articulated dynamics
model. We used the regressor form of the dynamic equation to get a least-square solution. Then the prediction
capability of the identified model is tested on a test trajectory, similar to the combined trajectories used for
identification, and a robot movement obtained using manual command via a master arm. Finally the capability
to detect perturbations is tested on movements where the robots contacted and pushed some objects. As
conclusion, we examine the capability to monitor for unpredicted behaviour such as collisions or friction with
an unknown environment. This capability is however limited to the case when the robot is not moving.
Furthermore, when using pressure-controlled servo-valves, the dynamical model will also detect internal
failure in the servo-actuators.
Corresponding Author:
BIDARD CATHERINE
catherine.bidard@cea.fr
Robotics & Interactive Systems Unit - CEA , BP6, F92265 Fontenay-aux-Roses Cedex, France
320
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-361
GENERIC CONTROL SYSTEM DESIGN FOR THE CASSETTE
MULTIFUNCTION MOVER AND OTHER ITER REMOTE HANDLING
EQUIPMENT
MICHAEL IRVING, J.PALMER (1) M.SIUKO (2)
(1) EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany (2)Tampere University of Technology, PO Box
589, 33101 Tampere, Finland
A fundamental difference between ‘robotic’ type control systems for normal industrial environments and those
intended for nuclear environments, is how they react to failures in the equipment they control. Normally,
troubleshooting and recovery from failure are carried out locally to the failed equipment, which for the remote
handling (RH) equipment used to carry out in-vessel ITER maintenance, is not possible due to the radiation
levels involved. Control systems for these environments need to be designed with fault detection, diagnosis
and recovery at the outset, since it is quite possible for a failure in the RH equipment to render it unrecoverable
and therefore compromise the whole project. Past experience in this field indicates that mechanical handling
equipment controllers are generally made by the company supplying the mechanics, and if as is likely in ITER,
different items of RH equipment are supplied by different manufacturers, each will possess a different type of
controller with different philosophies running different HMI’s, as has already happened. Across the entire suite
of RH equipment, this requires an enormous amount of technical knowledge to give effective support. The
alternative approach is to recognise that from the RH control perspective, equipment whose size, function and
purpose may be totally disparate, are likely to be drivable using virtually identical controllers. With RH
personnel needing in-depth knowledge of a now reduced range of equipment, incorporating modern reliability
and recovery techniques, it should be possible to detect failures before the task being carried out is
compromised, and certainly while recovery is still possible. Furthermore, controller subsystems designed and
built in modules will ease later upgrades, as well as making repairs simple and quick. This paper will explore
the design and operation of RH control systems specifically for ITER-type RH applications, with a view to
identifying guidelines to the equipment suppliers which will constrain the proliferation of controllers that
would otherwise naturally occur. This approach is particularly timely, since a new suite of RH equipment for
ITER is currently being specified, starting with the Cassette Multifunction Mover (CMM) to be used to
transport the divertor cassettes into and out of the vessel. Using this as an initial example, a generic type of
control system will be presented, which could be used as a basis for control of other ITER RH equipment.
Corresponding Author:
MICHAEL IRVING
mike.irving@brasimone.enea.it
Remote Maintenance Group (UTS Tecnologie Fisiche Avanzate), ENEA CR Brasimone, 40032 Camugnano
(BO), Italy
321
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-374
ANALYSES OF THE ITER VACUUM VESSEL WITH THE USE OF A
NEW MODELLING TECHNIQUE
ROZOV VLADIMIR, E. D'AGATA, K. IOKI, M. MORIMOTO, G. SANNAZZARO, R. TIVEY, YU. UTIN
Same as Corresponding Address
The on-going design development of the ITER Vacuum Vessel (VV) has been supported and accompanied
recently by extensive analytical studies of its different structural aspects. The use of a newly developed
parametric model of the ITER VV sector with port structures has enabled various assessments of this structure
in detail. Implementation of new features in the model and new approach for its generation eases updates and
modifications of the model, reducing a time necessary to introduce them into the model and allowing it to keep
pace with discussions on planned design change. It helps to keep the model up-to-date with the latest status of
the design and to minimise the time lag between the latest reference design documentation and the available
set of analyses validating it. A number of assessments have been accomplished recently for the ITER VV. An
assessment of its confinement function shows that it is able to fulfil its function and to keep the scale of any
mechanical failures limited to local areas even in the cases of a deflagration inside the chamber. An assessment
of the VV under hydrostatic test coolant pressure load has enabled an estimate to be made of its performance
as a double-wall structure pressurised in the interspace between the shells. Definition of the forces developed
in different regimes in the housings of all blanket supports which act as local links between the inner and the
outer shells has provided grounds for their further design improvement, correlated with the problem of the VV
manufacturing. Some analyses of the local reinforcements have been done since the new concept of the VV
support has been adopted. The implementation of complementary specialised tools for the automatic allocation
of the distributed masses has enabled a realistic representation to be made of the mass and inertia properties of
the real object and its parts in the program-generated model of the VV. This makes it possible to perform
assessments of the frequency and dynamic characteristics of the VV and other inertia-related issues. An
assessment of the triangular support region following recent design changes has been conducted using a
complementary model developed as a result of the implementation of the new modelling system for the
purpose of generation of local detailed models consisting of solid elements. The paper describes the main
results of these latest assessments of the ITER VV as part of the analytical support of the on-going design
work.
Corresponding Author:
ROZOV VLADIMIR
rozovv@itereu.de
ITER Joint Work Site, Boltzmannstr. 2, D-85748 Garching, Germany
322
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-389
ITER ARTICULATED INSPECTION ARM (AIA): GEOMETRIC
CALIBRATION ISSUES OF A LONG-REACH FLEXIBLE ROBOT.
D. ARHUR (1), Y. PERROT(1) C. BIDARD(1) J.P. FRICONNEAU(1) J.D. PALMER(2) C. TALARICO(3)
(1)Robotics and Interactive Systems Unit – CEA. BP6 F-92265 Fontenay aux Roses Cedex France (2)EFDA CSU Garching,
Boltzmannstrasse 2, 85748, Garching, Germany (3)ENEA –CR Frascati Fusion Technology Division – Via E. Fermi, 45
00044 Frascati Italy
This project takes place in the Remote Handling (RH) activities for the next step of the fusion reactor ITER.
Close inspection task of the Vacuum Vessel first wall of ITER with a long reach robot motivates
improvements on accuracy of the end effector’s position. The aim of the R&D program performed under
EFDA work program is to develop a flexible model of an IVP-like system. The first phase of the project
concerned the development of a IVP-like physical model. As the characteristics of the structure are the big
dimensions, the high number of joints with reduction of mass, significant compliance of the structure occurs.
However, geometric parameters but also non-geometric parameters such as stiffness must be taken into
account in the flexible model. The output of the model is the geometric configurations of the arm including the
end effector’s position. As our goal is to simulate the behaviour of a IVP-like system, a precise calibration of
the parameters is essential. The set of parameters is obtained using a non linear and multivariable optimisation
method. Its aim is to reduce the average distance between the end effector’s position stemming from the model
and the measured position by optimising a set of parameters. The identification method is first tested on
simulated positions to validate the feasibility of the approach and lastly applied to practical experimentation
performed on the IVP first module. On the practical experimentations, we show that taking into account the
flexibilities improves accuracy on the end effector’s position of the first module. The clearance on the rotation
axis and the repeatability of the mechanics affects the results of the optimisation because those phenomenons
are not predicted by the model. The mechanics must be improved to tend towards better results. Nevertheless,
the calibration method is more general, faster and improves the precision on the end effector’s position of the
first module. The extrapolation of the model to the whole arm will enable to compensate errors from more than
1 meter range in some configurations to a final accuracy of few centimetres. Once the whole arm will be
available, a global identification will have to be performed with the same method validated on the first
segment.This paper presents the physical model of the robot, the results stemming from the simulation and
measurements test campaign and at last the benefits of using this kind of model on the accuracy of the end
effector’s position.
Corresponding Author:
D. ARHUR (1)
delphine.arhur@cea.fr
Robotics and Interactive Systems Unit – CEA. BP6 F-92265 Fontenay aux Roses Cedex France
323
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-393
ITER ARTICULATED INSPECTION ARM (AIA) : R&D PROGRESS ON
VACUUM AND TEMPERATURE TECHNOLOGY FOR REMOTE
HANDLING.
YANN PERROT (1), J.J. CORDIER(2) J.P. FRICONNEAU(1) L. GARGIULO(2) E. MARTIN(3) J.D.
PALMER(4)
(1) DTSI - CEA BP6 92265 Fontenay aux Roses France (2) DRFC – CEA Cadarache, 13108 St Paul Lez Durance France (3)
EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany (4) ITER International Team, Boltzmannstrasse 2,
85748, Garching, Germany
This project takes place in the Remote Handling (RH) activities for the next step of the fusion reactor ITER.
The aim of the R&D program performed under EFDA workprogramme is to demonstrate the feasibility of
close inspection (e.g. for viewing and leak testing) of the Divertor cassettes and the Vacuum Vessel first wall
of ITER. We assumed that a long reach and limited payload carrier penetrates the first wall using the 6
penetrations evenly distributed around the machine designed for the In-Vessel Viewing System (IVVS). A first
phase of the project concerned the analysis of the requirements to perform a realistic operation inside the
Vacuum Vessel, a conceptual and detailed design of a manipulator called IVP (In Vessel Penetrator).A scale
one mock up was manufactured, focusing what is the most demanding and what requires to be demonstrated
(electro mechanics in air and at room temperature). This demonstrator of an IVP module (2 degrees of
freedom) was finally successfully tested and gave confidence to meet ITER requirements. In parallel, a
feasibility study of limited maintenance operation under vacuum and temperature with the IVP system was
performed. This study was completed and the possible applicable technologies were selected. Some of these
are directly suitable for the design of IVP under ITER’s vacuum and temperature requirements but some others
needed developments which were validated on proof of principles mock ups.The next step of the study is the
design, manufacture and testing of a vacuum and temperature hardened prototype called Articulated inspection
Arm (AIA) which is foreseen to be tested in real Tokamak conditions at Tore Supra. As well as demonstrating
the potential for the application of an AIA type device in ITER, this program will also serve to explore the
necessary robotic technologies applicable to ITER’s IVVS deployment system. This paper presents the whole
AIA robot concept, the results of the test campaign on the prototype vacuum and temperature demonstrator
module.
Corresponding Author:
YANN PERROT (1)
yann.perrot@cea.fr
Robotics and Interactive Systems Unit – CEA. BP6 F-92265 Fontenay aux Roses Cedex France
324
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-404
ASSESSMENT OF A COOPERATIVE MAINTENANCE SCHEME FOR
ITER DIVERTOR COOLING PIPE
FRICONNEAU JEAN-PIERRE, O. DAVID CEA - LIST J.P. MARTINS CEA LIST J.D. PALMER EFDA
CSU A. TESINI ITER NAKA
EFDA CSU Garching, Boltzmannstrasse 2, 85748, Garching, Germany ITER International Team, ITER Naka Joint Work
Site, 801-1, Muouyama, Naka-machi, Naka-gun, Iberaki-ken 311-0193, Japan
Divertor cassettes are components that require scheduled maintenance in the lifetime of the experimental
fusion reactor ITER. During maintenance phases, the cassettes are disconnected from the reactor frame and
moved outside the vessel for refurbishment in hot cells. Connection and disconnection of the cassette means
removal of the electrical connector and cutting or welding of the cassette cooling pipes. The ability for the
maintenance tool to cut or weld these cooling pipes therefore becomes an essential part of the maintenance
process. Previous studies of the maintenance task clearly identified operations that will need to be performed
remotely. They can be sorted into assembly operations: clamping of the tool, alignment or release of the
pipe… and into process operations: cutting, welding, non- destructive testing of the weld quality. The latest
evolutions of the ITER design focused on a reduction of the divertor cooling pipes from the initial 6” straight
pipes to 2.5” bent pipes. As a consequence of this new definition, the initial maintenance scheme where
maintenance is completely performed from the inside of the pipe by one carrier becomes obsolete. Reflections
are now converging on a cooperative work between a carrier sent into the pipe to perform part of the
maintenance operations and an orbital tool positioned by a slave manipulator mounted on the Cassette Toroidal
Mover (CTM) and used to perform general maintenance tasks in the divertor. While access to the cutting
location from the inside of the pipe was chosen during previous studies because the way was always clear,
access to the cooling pipes with a slave manipulator mounted on the CTM is a more challenging operation.
Definition of the reference maintenance scheme involves assessment of operational conditions such as:
feasibility of the operation, complexity of the environment, space to operate with the manipulator in relation to
its size and poor viewing conditions. Identification of the critical areas and zones where access to the pipe is
critical is made by means of digital mock-up analysis of the Vacuum Vessel divertor region. Extraction of a
reference working area is made after examination of all pipe ducts and cassettes arrangement. Simulations of
maintenance scenario take place in this reference area with a manipulator model based on a Maestro like
architecture. Accessibility towards the working area with a virtual cutting or welding orbital tool is then
checked and discussed.
Corresponding Author:
FRICONNEAU JEAN-PIERRE
jean-pierre.friconneau@cea.fr
Robotics and Interactive Systems Unit – CEA-LIST DTSI-SCRI. BP6 F-92265 Fontenay aux Roses Cedex
France
325
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-422
RF TESTS OF THE ELECTRICAL INSULATIONS FOR THE
TOROIDAL STRUCTURES OF RFX
SONATO PIERGIORGIO, A.MASIELLO, G.MELLA, C.TACCON
The new load assembly of RFX is characterized by a 3mm thick copper shell composed of two halves joined
on the outer equatorial plane by copper plates, which short-circuit the gap on the outer side, whereas on the
inner it remains electrically insulated. In the region of the single poloidal gap, two copper layers overlap each
other for about 23 toroidal degrees and the electrical insulation between them is secured by a 2mm PTFE layer.
The maximum toroidal loop voltage attainable with the magnetizing winding of RFX is 700V and in the case
of fast plasma current termination the observed peak reaches roughly the same value. The maximum poloidal
loop voltage is instead about one order of magnitude lower than the toroidal one. All the insulating gaps of the
shell need to be tested during the assembly phases of the torus, to verify the good conditions of the insulation.
In particular the poloidal gap on the shell shall be tested at least at 1kV. Since the toroidally shaped shell is a
single conducting structure with only one insulated gap in both direction (poloidal and toroidal), it is not
possible to apply an electric potential difference across the poloidal gap by means of a conventional DC or AC
generator. Different methods have been studied, based on a capacitor bank discharge, but the results showed
that the pulse duration and voltage level required for the test would impose the realization of an expensive
custom-made high voltage circuit. On the other hand it was found that, using a relatively small RF amplifier
(100 W) tuned at the natural frequency of resonance of the shell, sufficiently high voltage could be reached.
This innovative system was initially set up on the old RFX shell, which has approximately the same geometry
of the new one, to select the best matching equipment for the RF amplifier and to verify both the maximum
attainable potential difference at the poloidal gap and the ground effect. In fact it was found that the shell
behaves like a magnetic antenna and that the ground proximity reduces strongly the electromagnetic field, thus
also the voltage that can be applied to the poloidal gap. The RF system was then successfully used for the
acceptance test of the insulating poloidal gap of the shell at more than 1 kV, while 40W was the maximum
supplied RF power. These tests demonstrated that the RF method for applied-voltage tests is of general use and
it can be applied to closed conducting structures with a single insulating gap.
Corresponding Author:
SONATO PIERGIORGIO
piergiorgio.sonato@igi.cnr.it
Consorzio RFX, Corso Stati Uniti, 4 – I35127 Padova - Italy
326
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-435
OPERATIONAL EXPERIENCE FEEDBACK IN JET REMOTE
HANDLING
DAVID OLIVIER (1), A.B. LOVING (2) J.D. PALMER (3) S. CIATTAGLIA (3) J.P. FRICONNEAU (1)
(1)Robotics and Interactive Systems Unit CEA BP6 F-92265 Fontenay aux Roses Cedex France (2)UKAEA Fusion
Association, Culham Science Centre, Abingdon Oxfordshire, OX 14 3DB (3) Close Support Unit Garching –Boltzmannstr. 2,
D-85748 Garching Germany
The radiation levels expected in ITER during the latter stages of machine operation are such that maintenance
work cannot be carried out by human intervention. Remote Handling was therefore defined by the ITER
collaborators at the beginning of the project, as the nominal solution for the maintenance of the reactor. The
feedback provided by RH platforms developed for ITER within the L6 and L7 projects such as the Divertor
Test Platform (DTP) at Brasimone (Italy) and the In Vessel Transporter (IVT) at Naka (Japan) is clearly a step
further for the definition of ITER’s RH in several fields. But one has to admit that there is still a significant
amount of work to be done, especially considering the evolution in the machine design from ITER’98 to ITER
2001. JET is the only operating platform within fusion where RH techniques have been developed to a stage
that allows in-vessel maintenance work to be carried out fully remotely. JET’s RH team developed a
methodology and a rational approach that helped them to succeed in the task. JET’s experience clearly shows
that the gap between the first prototype and its upgrades to make it ready for operational use and perform
maintenance work in a safe and repeatable way has a manpower cost which is often under-estimated just like
the aspect of having a local team that you can rely on to solve any occurring problems and develop new
applications. As a simple rule of thumb, one can assume that following its original design and manufacture, the
effort required to properly prepare the RH equipment for real operations involved as much work as creating the
equipment in the first place. This paper presents some general rules that can be used to make the distinction
between the needs for hands-on activities and Remote Handling according to JET’s experience and ITER
needs and makes a summary of the experience gained by JET people during the development and operation of
their RH equipment which could be directly applied to ITER. Finally starting from the JET example, this
document tries to give ITER some references for the evaluation of the amount of work and of the manpower
cost that is really needed for a complete Remote Handling system to become fully operational and reliable.
Corresponding Author:
DAVID OLIVIER (1)
odavid@cea.fr
Robotics and Interactive Systems Unit – CEA BP6 F-92265 Fontenay aux Roses Cedex France
327
- G - Vessel-in vessel Engineering and Remote Handling.
P4T-G-509
IGNITOR PLASMA CHAMBER STRUCTURAL DESIGN WITH
DYNAMIC LOADS DUE TO PLASMA DISRUPTION EVENT
CUCCHIARO ANTONIO, BIANCHI ALDO (2), CRESCENZI CLAUDIO (1), LINARI MAURO (3), LUCCA FLAVIO
(3), MARIN ANNA (3), MAZZONE GIUSEPPE (1), PARODI BRUNO (2), PIZZUTO ALDO (1), RAMOGIDA
GIUSEPPE (1), ROCCELLA MASSIMO (1), PROF. COPPI BRUNO (4)
(1) Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy (2) Ansaldo Ricerche, Corso Perrone
25, 16152 Genova (GE), Italy (3) L. T. Calcoli, Piazza Prinetti 26/B, 23805 Merate (LC), Italy (4) MIT, 02139 Cambridge
(Ma), USA
The new reference plasma disruption for IGNITOR produces a significant increase of electromagnetic (EM)
loads requiring a dynamic elastic-plastic structural analysis of the IGNITOR plasma chamber (PC). The EM
loads due to the worst disruption event (VDE) was calculated using the MAXFEA 2D code. The uniform 26
mm PC thickness (as envisaged in the previous design) did not comply with the new E.M. requirements in
term of stresses and deformation. New Design Plasma Chamber wall thickness distribution has been defined
approaching a step by step optimization with the aim to minimize the vertical displacement complying with the
allowable plastic strains. The analysis is non-linear, due to boundary conductions and material properties, and
it’s necessary to modelling the entire (360 ) PC structure because the various load components are distributed
with different poloidal periodicity. As result the IGNITOR PC is capable to withstand, according to the ASME
III code rules, several hundred of cycles under plasma disruption conditions. The main results of the analysis
demonstrate that the plastic deformation is below the ASME limits and the maximum permanent displacement
is limited to few millimeters.
Corresponding Author:
CUCCHIARO ANTONIO
cucchiaro@frascati.enea.it
Associazione ENEA-EURATOM sulla Fusione, C.P. 65, 00044 Frascati (RM), Italy
328
- H - FUEL CYCLE.
P1C-H-17
ADVANCED PROCEDURES FOR TWO-STAGE REPETITIVE PELLET
INJECTOR.
PAVARIN DANIELE, FRANCESCONI ALESSANDRO NIERO FEDERICO RONDINI DAVIDE ANGRILLI
FRANCESCO
Centero of Studies and Activities for Space (CISAS G.Colmbo) University of Padua Via Venezia 15 35131 Padova Italy
Two stage pellet injector for fusion experiments are powerful machine to accelerate refuelling pellet at very
high velocity ensuring penetration in the plasma core also through very dense and energetic plasmas. Besides
the possibility of accelerating pellet at very high velocity they are effected by several problems as: gas
following the pellet, which may contaminate the plasma, projectile strength and system reliability. CISAS has
developed a two-stage unit based on the concept of fusion injector which is able to accelerate plastic projectile
weighting 100 mg at 5.5 km/s. The achievement of this goal required the set-up of advanced control and
diagnostic procedures and new technology solutions which may become useful for the development of the
future fusion pellet injectors. Particularly CISAS Gun implements high release pressure check valve which
could be useful to reduce the gas following the projectile, diagnostic procedure are applied to check the gun
status with no needs of disassembling the unit, and finally active piston techniques are applied to control the
shape of the pressure profile behind the projectile. In the paper a detailed description of the procedure set-up in
the CISAS facility is presented and a possible application to future fusion experiments discussed.
Corresponding Author:
PAVARIN DANIELE
daniele.pavarin@unipd.it
CISAS University of Padua Via Venezia 15 35131 Padova Italy
329
- H - Fuel Cycle.
P1C-H-38
STUDIES OF PELLET DELIVERY AND SURVIVABILITY THROUGH
CURVED GUIDE TUBES FOR FUSION FUELING AND IMPLICATIONS
FOR ITER
COMBS, STEPHEN, L. R. BAYLOR, C. R. FOUST, T. C. JERNIGAN, AND D. A. RASMUSSEN
Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6169
Injection of solid hydrogen pellets will be the primary technique for depositing fuel particles into the core of
ITER burning plasmas. Gas fueling is calculated to have much lower fueling efficiency in ITER than present
day experiments, thus pellet fueling will be particularly important. For many years, pellets have been injected
from the outside mid-plane of experimental fusion devices (magnetic low-field side for tokamaks), and
researchers strived for high pellet speeds (up to ~4 km/s) to achieve deep pellet penetration and better fueling
efficiency [1]. In the past several years, pellet technology development and experiments have concentrated on
pellet fueling from the magnetic high-field side for tokamaks [2], with significantly deeper mass deposition
and improved fueling efficiency observed at relatively low pellet speeds (<300 m/s). These injection schemes
require the use of curved guide tubes to route the pellets from the acceleration devices to the inside wall launch
or vertical launch locations; and thus, the speed must be limited to ensure pellet survivability. To more
thoroughly understand and document the capability of curved guide tubes for the alternative injection schemes,
experimental studies have been carried out at the Oak Ridge National Laboratory (ORNL), including mock-up
tests of guide tube installations for DIII-D, JET, LHD, and FIRE. For the actual installations on the machines
and fueling experiments, the speed limits have proved to be in good agreement with the results from the ORNL
mock-up tests. For these installations, the speeds of deuterium pellets must be limited to a few hundreds of
meters per second for reliable delivery of intact pellets. Presently, an experimental simulation of the ITER
guide tube installation for inside pellet launch is being set up in the lab and will be tested with 3-mm
deuterium pellets. The test results from the previous mock-ups will be summarized in the paper, and the new
data from the ITER mock-up will be presented and discussed. The implications of these results for ITER pellet
fueling will also be discussed. [1] A. Géraud et al., Proc. 20th Symposium on Fusion Technology (1998) 941–
944. [2] P.T. Lang et al., Phys. Rev. Lett. 79 (1997) 1487–1490.
Corresponding Author:
COMBS, STEPHEN
combssk@ornl.gov
Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6169
330
- H - Fuel Cycle.
P1C-H-93
PELLET INJECTORS FOR STEADY STATE FUELLING
VINYAR IGOR, A.GÉRAUD(2), H.YAMADA(3), A.LUKIN(1), R.SAKAMOTO(3), S.SKOBLIKOV(1), A.UMOV(1)
Y.ODA(4), G.GROS(2), I.KRASILNIKOV(1), P.REZNICHENKO(1), V.PANCHENKO(1)
(1)PELIN Laboratory,20a,Berezhkovskaya Emb.,Moscow,123995,Russia (2)Euratom-CEA,CEA/Cadarache,F-13108,St-Paul
Lez Durance,France (3)National Institute for Fusion Science,322-6,Toki,509-5292,Japan (4)Mitsubishi Heavy
Industries,Kobe,652-8585, Japan
Selected also for oral presentation
O1B-H-93
Current successful operation of TORE SUPRA and LHD, as well as ITER in the future, should be supported
by pellet injection in order to produce high performance plasmas with discharge durations up to 1000s. This
paper presents pneumatic pellet injectors and its implementation on the LHD and TORE SUPRA, and a new
centrifuge injector for steady state fuelling. All pellet injectors are based on the screw extruder technology
developed by the PELIN Laboratory. This feed system, well suited for continuous operation, is coupled to a
pellet cutting system forming a steady state pellet generator. Two conventional launchers are used for pellet
acceleration: pneumatic gun and centrifuge. A distinguishing feature of the LHD pellet injector is a cooling
system based on two cryorefrigerators (3 W at 4.2 K in total). The pellet injection is available 4 hours after the
start of cooling down from the room temperature. Hydrogen is continuously pushed out from the extruder in
the shape of cylindrical ice rod of 2 or 2.5 mm diameter with the maximum rate of 15 mg/s. Pellets can be cut
off from the rod and ejected at frequency up to 11 Hz and velocities from 150 to 650 m/s with a reliability
better than 99%. The amount of gas is suppressed to less than 1.5 Pam3 per one pellet launch. The injector has
been employed for LHD plasma fuelling and demonstrated quasi-steady state high density operation (8 x
1019m-3) for 2 s. This duration is limited by heating capability not by pellet injection. The TORE SUPRA
pellet injector was equipped with a pellet size regulator for real-time fuelling adjustment during the plasma
discharge using feedback control data. Pellet diameters are 1.7 mm, but pellet lengths can be changed from 1.5
to 3.5 mm. Operational injector characteristics in steady state mode are: (2-6)1020 at/pellet, 120-700 m/s
velocity, 0-10 Hz frequency, 98% reliability. High field side, vertical and low field side injections were
performed in plasma experiments in 2003-2004. A distinguishing feature of the centrifuge injector is a curved
barrel for pellet acceleration whose entrance section is aligned with the rotational axis of the rotor. So, the
barrel entrance section, into which a pellet is fed, does not replace relatively to the fuel rod extruded from a
screw extruder. This design allows pellets to be fed in the barrel entrance section in any moment. This injector
is under testing now.
Corresponding Author:
VINYAR IGOR
pelin@delfa.net
PELIN Laboratory, Ltd., 20a, Berzhkovskaya Emb., Moscow, 123995, Russia
331
- H - Fuel Cycle.
P1C-H-107
JET CONTRIBUTIONS TO THE ITER FUEL CYCLE ISSUES.
C. GRISOLIA,
Since 2000, the JET fusion machine is operated in the frame of European Fusion Development Agreement. It
is the unique world fusion device licensed for operating with tritium fuelling and beryllium. A dedicated task
force on fusion technology (TF-FT) aims at answering to ITER relevant technological issues making use of
JET facilities and of the related operating experience. In the frame of tritium in the tokamak, surface analyses
allow an estimation of Be and C deposition on plasma facing components. According to the obtained results,
Be is transported towards the upper tiles of the inner divertor where it is stacked. Carbon, after deposition, is
re-eroded through chemical sputtering and transported towards the inner flat tiles. Improved calibrations
indicate that the total Be deposited in the inner divertor is of the order to 35-40 g leading to total carbon influx
of 570 to 870 g. This value fits by less than a factor two to the spectroscopic evaluation. Surface analyses
dedicated to 13C characterisation just after 13CH4 puffing plasma experiments show that 40% of the 13C is
trapped in the inner divertor tiles confirming the flow in the scrape off layer. Estimation of ITER in-vessel
tritium retention have shown that overnight in situ detritiation is needed during operation. Detritiation
processes based on laser or flash lamp are being investigated for JET configuration. According to laser studies,
ablation energy threshold is of 0.4 J/cm2 for a co-deposited layer whereas is of 1J/cm2 for graphite surfaces. A
removal rate of 1 m2/hour is obtained for 20 µm co-deposited layer from TEXTROR and Tore Supra. Flash
lamp in situ tests are scheduled in April 2004 to confirm ex-situ results (cleaning rate > 3m2/hour for
50microns). On gas exhaust control, the JET Active Gas Handling System has been used to test a prototype
cryosorption panels during Trace Tritium Experiment. Results will be addressed in the paper. In the frame of
the design project of a water detritiation facility for JET, key components for such a system are being studied.
The most promising catalyst/packing mixtures for the Liquid Phase Catalytic Exchange (LPCE) column are
currently tested in endurance tests. After recalling briefly the TF-FT activities and especially those devoted to
ITER licensing involving Plasma Facing Components, Tritium processes and safety, this paper will be focused
on the presentation of a comprehensive picture of the Tritium fuel cycle in a fusion facility.
Corresponding Author:
C. GRISOLIA
christian.grisolia@cea.fr
Assoc. Euratom-CEA sur la Fusion, CEA Cadarache, DSM/DRFC/STEP, 13108 Saint Paul Lez Durance, France
332
- H - Fuel Cycle.
P1C-H-153
COMPARISON OF MODELLING OF TRITIUM RELEASE FROM
CERAMIC BREEDER MATERIALS
MUNAKATA KENZO, YOKOYAMA YOSHIHIRO (1) PENZHORN R. -D. (2) OYAIDZU MAKOTO (3) OKUNO
KENJI (3)
(1) Kyushu University, Kasuga 816-8580, Japan (2) Research Center Karlsruhe, Tritium Laboratory Karlsruhe, 76021
Karlsruhe, Germany (3) Shizuoka University, Radiochemistry Research Laboratory, Faculty of Science, Szuoka, 422-8529,
Japan
In most current designs of D-T fusion reactor blankets employing ceramic breeder materials, the use of a
helium sweep gas containing 0.1 % of hydrogen is contemplated to extract tritium efficiently via isotopic
exchange reactions. However, at lower temperatures, the release process of tritium from the breeders is
dominated by the desorption of tritiated water and is therefore rather slow since the rate of isotope exchange
reactions is considerably low. With this background, the authors investigated the effect of water vapor on the
releases of tritium from a lithium silicate ceramic breeder material. Out of pile tritium release experiments
were conducted using the ceramic breeder irradiated in a research reactor. Tritium was released from the
lithium silicate breeder material using a nitrogen sweep gas with 0.1 % of water vapor. As a result, it was
found that the addition of water vapor to the sweep gas greatly enhances the release rate of tritium from the
ceramic breeder. These are probably caused by the acceleration of the exchange reaction that takes place on
the surface of the breeder material. The result of tritium release experiment with the wet sweep gas was
analyzed using several numerical models. The surface effect was eliminated in the model, since the surface
reactions were thought being significantly enhanced by the wet gas purge. The results of the analysis indicate
that simple linear driving force model or diffusion model cannot reproduce the experimental tritium release
curve. Thus, the trapping of tritium caused by trapping sites was also considered in the model. Then, it was
found that both of the models with the trapping site effect well reproduce the experimental tritium release
curve. Moreover, the authors tested another model in which a resistance between the bulk crystal phase and the
surface of the crystal grain was considered. It is known that the disordered layer close to the surface becomes a
resistance to the migration of hydrogen in metals. As a result, this model was also found to reproduce the
experimental tritium release curve. In the presentation, the details of the model were explained, and the
reproducibility of the experimental result was compared.
Corresponding Author:
MUNAKATA KENZO
kenzo@nucl.kyushu-u.ac.jp
Kyushu University, Interdisciplinary Graduate School of Engineering Science, Kasuga 816-8580, Japan
333
- H - Fuel Cycle.
P1C-H-217
ASSESSMENT OF THE ITER DWELL EVACUATION
WYKES MICHAEL, FEDERICI GIANFRANCO
ITER International Team, Boltzmannstrasse 2, D-85748 Garching, Germany
During the “dwell” period between ITER plasma current pulses the pressure in the plasma chamber has to be
pumped down to a level of ~ 0.1 mPa to allow for an orderly pre-fill and breakdown to initiate the subsequent
plasma current pulse. The shortest reference evacuation time corresponds to the dwell between successive
ITER 400 s burn pulses, each dwell being 1400 s long. The dwell evacuation is dominated by the outgassing
from the plasma-facing components, the majority of which are armoured with beryllium (~700 m2), with a
minority in CFC on the lower divertor vertical targets (~50 m2). During plasma discharges, impinging
deuterons and tritons load the implantation layer of the armour to near-saturation conditions. From previous
experimental measurements and theoretical studies, it is known that the outgassing rate decays from an initial
value that depends on the armour material and temperature, and the energy and implantation time of the
incident particles, according to a power law in elapsed time (t-n), the exponent n being ~0.7. During the
subsequent dwell evacuation, the implanted atoms are desorbed and constitute the main load on the primary
torus cryo-sorption pumps, particularly during the latter stage of evacuation when the pressure is low. The
results of parametric studies are reported to assess the effect of various factors that may affect the outgassing
rate (e.g., mixing of material, effect of temperature). These latter results delineate the domain in which the
terminal pressure is acceptable and indicate the amount of additional pumping that will be provided by the
neutral beam cryo-sorption pumps in order to attain the required dwell terminal pressure under the most
adverse, but realistic, outgassing characteristics.
Corresponding Author:
WYKES MICHAEL
wykesm@itereu.de
ITER International Team, Boltzmannstrasse 2, D-85748 Garching, Germany
334
- H - Fuel Cycle.
P1C-H-247
STRATEGY FOR DETERMINATION OF ITER IN-VESSEL TRITIUM
INVENTORY
MURDOCH, DAVID, CRISTESCU, IOANA-RUXANDRA (1); LÄSSER RAINER (2)
(1) TLK, Forschungszentrum Karlsruhe, Postfach 3640, D76021 Karlsruhe, Germany. (2) EFDA Garching CSU, Max Planck
IPP, Boltzmannstr. 2, D-85748 Garching, Germany
Tracking of tritium inventory transfers on ITER will be essential to ensure that the safety limits established for
the mobilizable tritium inventory in the vacuum vessel (VV) are not violated. The large and highly variable
fuelling and exhaust flow rates, and the complex chemical composition of the tokamak exhaust stream, will
compromise the precision of direct flow rate and composition measurements. Thus a complementary
procedure to derive the in-vessel inventory at regular intervals is required, and this is described in the paper. In
ITER, the fuel cycle, the VV and the hot cell building are within the same tritium Material Balance Area
(MBA). The global tritium inventory of this MBA at any selected time can be derived from the previous
determination by measuring all tritium quantities entering and leaving the MBA and calculating the quantities
of tritium created (by breeding) and destroyed (by fusion reactions and decay) within it. This depends on the
fact that reliable measurements and/or calculations of each of these source terms can be made, and the methods
proposed for this will be outlined in the paper. At predetermined intervals the tritiated gases in all systems of
the fuel cycle will be transferred to the Storage and Delivery System (SDS), and tritium quantities measured
using in-bed calorimetry, to give the total inventory of mobile tritium in the fuel cycle system by addition. The
calculated global inventory of the MBA less the measurable fuel cycle inventory represents the trapped tritium
inventory, both inside and outside the VV. The ex-vessel portion of tritium bound in materials such as
catalysts, molecular sieves, pump oils, and component walls will be determined by building up a
comprehensive experimental data base on the evolution of tritium content in these materials from the start of
operations, enabling the in-VV portion to be derived by difference. The procedure for transfer of the
mobilizable gases to the SDS, assay of the tritium and redistribution of the gases before restart will be outlined
and estimates of the time required and the achievable accuracy of the determinations presented. Although the
VV and the hot cell building are within the same MBA, tritium inventories in activated components, PFC
flakes and dust transferred to the Hot Cell, and in refurbished components returned to the VV, will be assessed
in order to adjust the in-vessel inventory. Proposed methods for achieving this will be discussed.
Corresponding Author:
MURDOCH, DAVID
david.murdoch@tech.efda.org
EFDA Garching CSU, Max Planck IPP, Boltzmannstr. 2, D-85748 Garching, Germany
335
- H - Fuel Cycle.
P1C-H-275
REQUIREMENTS AND SELECTION CRITERIA FOR THE
MECHANICAL PUMPS FOR THE ITER TRITIUM PLANT
C J CALDWELL-NICHOLS, M GLUGLA (1), S WELTE (1), D MURDOCH (2)
(1) Tritium Laboratory, Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe, Germany (2) EFDA CSU, MPI,
Boltzmannstraße 2, D-85748 Garching, Germany
The tritium plant for ITER requires many mechanical gas pumps for its operation. The correct selection of
pumps is important not only to meet the requirements of the various processes in terms of pressures, flow rates
etc. but also must be satisfy more nebulous criteria of reliability and maintainability. In common with many
ITER components, they will not be purchased for several years but will then be required to operate over the 20
years of the operational lifetime of ITER and possibly beyond this time. It must be expected that there will be
failures over the lifetime of the tritium plant, so attention must be paid to the methods for maintenance, noting
that all components will be installed in glove boxes or other enclosures. A good prediction of the availability
of replacement pumps or components over this period will have to be made. An analysis of the requirements of
all pumping stages within the tritium plant will be made and these will be compared with performance of the
available pump types. Factors to be considered are whether to use single compound pumps in certain
applications or several individual smaller and simpler pumps. This analysis may lead to the reduction of the
types of pumps used. The types of pump motors will also be examined , noting that brushed motors are not
acceptable inside glove boxes due to the creation of carbon dust. The results of the analyses and tests of
candidate pumps in representative conditions (inclusion of filters, high Reynold’s number flow, long pipe
lengths) will be presented.
Corresponding Author:
C J CALDWELL-NICHOLS
chris.caldwell-nichols@hvt.fzk.de
Tritium Laboratory, Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe, Germany
336
- H - Fuel Cycle.
P1C-H-303
HIGH-POWER PULSED FLASHLAMP CLEANING OF CO-DEPOSITED
HYDROCARBON FILMS FROM PLASMA FACING COMPONENTS
K.J. GIBSON, G.F. COUNSELL (2) M.J. FORREST (2) M.J. KAY
(2) Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon,, Oxon.OX14 3DB, UK
The use of carbon-based materials for first wall components in the divertor region of tokamaks has been
observed to result in the formation of significant amorphous hydrocarbon deposits on both plasma facing
components and sub-divertor pumping assemblies. Such deposits, which take the form of thin films of varying
morphology, could lead to a high rate of in-vessel tritium retention in future fusion devices. Laboratory based
experiments at UMIST have demonstrated that photonic cleaning can provide a clean, efficient and rapid
method for removing such hydrocarbon films and represents a good candidate technology for the periodic
removal of co-deposits in ITER. The use of high power Xenon flashlamps as a source for this cleaning has
been demonstrated in air, inert atmosphere and vacuum, with effective film removal occurring at a fluence
threshold of between 1.9 and 2.5 J/cm2. The by-products of the cleaning process, both particulates and gases,
have been characterised using particle sizing spectrometry and quadrupole mass spectrometry respectively. It
is found that hydrogen, methane, acetylene, ethylene, ethane and carbon dioxide are the principal gaseous
products produced during the cleaning process, which also produces a significant fraction of particulates in the
size range 2-20mm. In an extension to this work, a series of experiments are described in which co-deposits on
divertor tiles from the MAST and Asdex Upgrade tokamaks have been removed using a newly assembled
flashlamp source. This source, capable of delivering instantaneous powers of up to 1GW/m2 and with pulse
duration of approximately 100ms, allows a more complete study of the scaling of the removal process at
higher fluences (up to 10J/cm2, well above the threshold) as well as a comparison of the relative importance to
the cleaning efficiency of the UV and visible components of the flashlamp spectral output. Finally we outline
plans for in-situ, full scale demonstrations of flashlamp removal of co-deposited films from the JET and ToreSupra tokamaks. This work was jointly funded by the UK Engineering and Physical Sciences Research
Council and EURATOM. The authors would also like to thank Dr Rudolf Neu of the Max-Plank Institut für
PlasmaPhysik, Garching for the loan of Asdex Upgrade tiles.
Corresponding Author:
K.J. GIBSON
k.gibson@umist.ac.uk
Department of Physics, UMIST, Manchester M60 1QD, UK
337
- H - Fuel Cycle.
P1C-H-358
GAS PUFFING BY MOLECULAR BEAM INJECTION IN ADITYA
TOKAMAK
S. B. BHATT, AJAI KUMAR, K. P. SUBRAMANIAN*, P. K. ATREY, C. V. S. RAO AND ADITYA TEAM
Institute for Plasma Research, Bhat, Gandhinagar-382428, Gujarat, India * Physical Research Laboratory, Ahmedabad –
380 009, INDIA
Plasma density control is a prime requirement of tokamak plasma. Various methods are used for fuelling the
gas for tokamak plasma. In normal gas puffing system, there is large angular distribution/ velocity profile in
injected gas molecules. Due to this, considerable number of rundown gas molecules are adsorbed on the
surface of wall and limiter. These adsorbed gas molecules are released from the surface during the plasma
discharge and causing more recycling of fuel gas and some time control of density is difficult. A new method
of gas fuelling in tokamak with molecular beam injection is developed for fast gas puff during the plasma
discharge. The molecular beam of fuel gas is formed by the expansion of the high-pressure gas through the
nozzle and skimmer. The penetration depth injected beam is more than normal gas puffing due to its energy.
Due more penetration and very less divergence of beam, 1.wall loading due to fuel gas reduces causing
reduction in recycling, and 2, there is an increase in the fuelling efficiency. A molecular beam injection system
for Aditya tokamak is developed in house by modifying a Piezo-electric gas inlet valve and using100 µm
diameter nozzle at 2 to 5 kg/cm2 gas pressure. This valve is mounted directly on tokamak. Series of
experiments are performed with this molecular beam system by puffing hydrogen gas of different pulse
duration, different number of pulses at various time of tokamak discharge. It is observed that there is increase,
in plasma density more than 1.5 times, bolometer signal, soft X-ray and reduction in H? signal. This paper
presents the results of molecular beam injection during discharge.
Corresponding Author:
S. B. BHATT
sbbhatt@ipr.res.in
Institute for Plasma Research, Bhat, Gandhinagar-382428, Gujarat, India
338
- H - Fuel Cycle.
P1C-H-438
INFLUENCE OF DEUTERIUM ON THE DESIGN OF THE JET WATER
DETRITIATION SYSTEM
CRISTESCU ION, I-R CRISTESCU 1, L. DÖRR 1, M. GLUGLA 1, A .BELL 2, D. BRENNAN 2, D. MURDOCH 3
1 Forschungszentrum Karlsruhe, Tritium Laboratory, Germany 2 JET, Abigdon, UK 3 EFDA CSU Garching
The development of a Water Detritiation System (WDS), i.e. configuration, performance testing of critical
components and system design is essential for both JET and ITER. The WDS at JET is foreseen to process the
tritiated water accumulated during operations and generated during decommissioning. The WDS for JET will
be based on the Combined Electrolysis Catalytic Exchange (CECE) process employing a Liquid Phase
Catalytic Exchange (LPCE) column. A direct combination of the WDS with a Cryogenic Distillation (CD) unit
is also under planning. The final goal is to convert tritiated water to tritium and deuterium enriched molecular
hydrogen and to further enrich tritium by cryogenic distillation of the hydrogen-deuterium –tritium mixture,
followed by recovery of pure tritium in the GC based isotope separation system of the AGHS of JET. A
decontamination factor of 104 for tritium is required along the striping section of the LPCE column in order to
discharge and essentially tritium free molecular hydrogen isotope product into the environment. The process
for the JET WDS was preliminary evaluated considering the deuterium concentration in the tritiated water to
be at natural level. A detailed analysis of the direct combination of the CECE-CD processes revealed the
necessity to consider also the deuterium content in the water to be processed. Therefore eleven samples from
different drums with tritiated water at JET have been measured for their deuterium concentration, which was
found to be in the range of 0.2-0.5% atomic ratio D/ (H+D+T). The presence of deuterium in tritiated water to
be processed will change the tritium distribution on molecular species. Therefore instead of two molecular
species, such as H2 and HT when deuterium is negligible, in the CECE and CD processes five molecular
species such as H2, HT, D2, HD, DT has to be considered. The constant equilibrium and separation factor for
CECE and CD processes are very different from one molecular specie to another which means that the length
of the LPCE and CD columns depends upon the molecular species which contain tritium in order to achieve a
required decontamination factor of tritium. Based on these measurements the interface between the CECE and
CD was evaluated in detailed and the optimum values for deuterium and tritium composition at this interface
were established.
Corresponding Author:
CRISTESCU ION
ion.cristescu@hvt.fzk.de
Forschungszentrum Karlsruhe GmbH, Postfach 36 40, 76021 Karlsruhe
339
- H - Fuel Cycle.
P1C-H-441
EXPERIMENTAL VALIDATION OF A METHOD FOR PERFORMANCE
MONITORING OF THE FRONT-END PERMEATORS IN THE TEP
SYSTEM OF ITER
B. BORNSCHEIN, M. GLUGLA (1) K. GUENTHER (1) T.L. LE (1) K.H. SIMON (1)
(1) Forschungszentrum Karlsruhe, TLK, P.O. Box 3640, D76021 Karlsruhe, Germany
The Tokamak Exhaust Processing (TEP) system within the Tritium Plant of ITER need to be designed such
that tritium is recovered from all exhaust gases produced during different modes and operational conditions of
the vacuum vessel. The reference process for the TEP system of ITER is called CAPER and comprises three
different, consecutive steps to recover hydrogen isotopes at highest purity for direct transfer to the cryogenic
Isotope Separation system. A tritium removal efficiency of about 1E+8 is required for TEP and is regularly
achieved in experiments with the semi-technical facility CAPER at the Tritium Laboratory Karlsruhe (TLK).
Expressed in terms of tritium concentrations the decontamination required by TEP corresponds to an outlet
concentration of about 1E-4 g/m^3 (1 Ci/m^3). The CAPER process developed at the TLK employs a Pd/Ag
permeator battery as the 1st step to separate more than 95% of the un-burnt DT fuel from impurities like
helium, hydrocarbons and water. These so called front-end permeators have to cope with a feed flow rate of
about 80 mol/h per 1 m^2 effective surface area. They have to be operated under conditions that avoid coking
of the permeation membranes by hydrocarbon cracking, since this process would lead to a reduction of the
effective surface area and therefore to a reduction of the performance of the permeator. In a series of tritium
experiments with the CAPER facility at TLK a method has been developed to determine the actual hydrogen
isotope permeability of the front-end permeator. During this experimental campaign the permeator has been
operated with DT (typically 50%T) mixed with tritiated methane under conditions that promote coking by
hydrocarbon cracking. The reduction of the permeator performance has then been determined by
measurements of so-called break-through curves. The front end permeator could all the times be successfully
regenerated after coking the membranes. The experimental results will be presented and the feasibility of
performance monitoring of the ITER front-end permeators will be described. Details of the regeneration
process will be reported and possible consequences for the design of the TEP system will be discussed.
Corresponding Author:
B. BORNSCHEIN
beate.bornschein@hvt.fzk.de
Forschungszentrum Karlsruhe, TLK, P.O. Box 3640, D76021 Karlsruhe, Germany
340
- H - Fuel Cycle.
P1C-H-461
PROTECTION OF THE PRIMARY CIRCUITS AND EFFECT ON THE
DESIGN OF THE INNER DEUTERIUM / TRITIUM FUEL CYCLE OF
ITER
M. GLUGLA, C. CALDWELL-NICHOLS (1), I.R. CRISTESCU (1), L. DOERR (1), G. HELLRIEGEL (1), D.
MURDOCH (2), P. SCHAEFER (1)
(1)Forschungszentrum Karlsruhe, Tritium Laboratory, PO Box 3640, D 76021 Karlsruhe, Germany (2)EFDA CSU, MPI fuer
Plasmaphysik, Boltzmannstr. 2, D 85748 Garching, Germany
The inner deuterium / tritium fuel cycle of ITER comprises different, but strongly interlinked subsystems,
namely the Tokamak Exhaust Processing (TEP) system, the Isotope Separation System (ISS) and the Water
Detritiation System (WDS), the Storage and Delivery System (SDS) together with the Long Term Storage
(LTS) and the Analytical System (ANS). Detailed Process Flow Diagrams (PFD’s) and even Pipe and
Instrumentation Diagrams (P&ID’s) have been prepared and were included in the Final Design Report of ITER
(2001). For each of the subsystems an initial Failure Mode and Effect Analysis (FMEA) was carried out
individually. An Outline Flow Diagram for the inner fuel cycle was prepared for design integration and
eventually an FMEA considering the inner fuel cycle as a whole became available. Confinement of tritium is
certainly the ultimate safety goal. However, the protection of components such as sensors, pumps or vessels
against over-pressure or over-temperature is of great concern, even at levels significantly below values at
which the sensors or components would loose their mechanical integrity. The design shall take into account the
necessity to validate and test the protection measures, noting the contamination of the primary system with
tritium and the restricted access due to secondary containments. In view of primary safety, the subsystems of
the ITER fuel cycle have been designed on the basis of the safety philosophy established and practiced at the
Tritium Laboratory Karlsruhe (TLK). However, modern international standards for functional safety
management such as the IEC 61508 are now available and should be evaluated for applicability in the ITER
Tritium Plant. The general philosophy and the principles for over-pressure and over-temperature protection
within the design of the inner fuel cycle of ITER will be explained in detail. Examples will be presented for
selected subsystems.
Corresponding Author:
M. GLUGLA
glugla@hvt.fzk.de
Forschungszentrum Karlsruhe, Tritium Laboratory, PO Box 3640, D 76021 Karlsruhe, Germany
341
- H - Fuel Cycle.
P1C-H-467
EVALUATION OF SUPER CRITICAL HELIUM AS A COOLANT FOR
DIII-D TYPE CRYOCONDENSATION
BAXI, C.B.,
DIII–D tokamak uses three cryocondensation pumps for plasma density control. Each DIII–D pump consists of
a series of concentric stainless steel tubes. The pumping surface is a 10 m long 25 mm diameter stainless steel
tube. The pumping surface of each of the three cryocondensation pumps is about 1 sq m in area and is
maintained below 5 K by cooling with a two phase helium (1.3 atm, 4.35 K). The two-phase helium (TP) was
chosen for DIII-D because it is available on DIII-D site and is used for NB and other applications. The
pumping speed is about 30000 l/s per pump. The three pumps inside DIII-D have performed as expected for
last several years. Super conducting machines under construction such as KSTAR and SST-1 have
supercritical (SC) helium available on site and would prefer to use it for cooling the cryocondensation pumps.
The typical condition of the available helium is 0.4 MPa (3.94 atm) pressure and 4.2 K temperature. The
design of DIII-D cryocondensation pump is simple, robust, inexpensive and reliable. This study was
undertaken to evaluate if the supercritical helium can be used as a coolant for GA design of the
cryocondensation pump. Thermodynamic, thermal hydaulic and stability evaluation was done. It is concluded
that with super critical helium a flow rate of 50 to 60 g/s (compared to 5 to 10 gm/s with two phase helium)
will be required to achieve a similar performance. The co-axial insert used in DIII–D helium panel will not be
required with SC helium.
Corresponding Author:
BAXI, C.B.
baxi@fusion.gat.com
General Atomics, P.O. Box 85608, San Diego, California 92186-5608
342
- I - MATERIALS TECHNOLOGY AND BREEDING BLANKETS.
P1C-I-1
USE OF THE SPIRAL 2 FACILITY FOR MATERIAL IRRADIATIONS
WITH 14 MEV ENERGY NEUTRONS
MOSNIER ALBAN, R. ANNE (1) Y. HUGUET (1) X. LEDOUX (4) M. LIPA (2) PH. MAGAUD (2) G.
MARBACH (2) F. PELLEMOINE (1) D. RIDIKAS (3) M.G. SAINT-LAURENT (1) A.C.C. VILLARI (1)
(1) GANIL, BP 55027, 14076 Caen, France (2) CEA/DSM/DRFC, CEA/Cadarache, 13108 Saint Paul Lez Durance, France
(3) CEA/DSM/DAPNIA, CEA-Saclay, 91191 Gif-sur-Yvette, France (4) CEA/DAM/DPTA, BP 12, 91680, Bruyeres-le-Chatel,
France
The primary goal of an irradiation facility for fusion applications will be to generate a material irradiation
database for the design, construction, licensing, and safe operation of a fusion demonstration reactor (e.g.,
DEMO). This will be achieved through testing and qualifying material performance under neutron irradiation
that simulates service up to the full lifetime anticipated in the demonstration fusion reactors. Preliminary
investigations of 14 MeV neutron effects on different kinds of fusion material could be assessed by the
SPIRAL 2 project at GANIL (Caen-France) with first beams expected by 2009. This would allow to prepare
and validate experiences which are scheduled in IFMIF during 2010-2015. It concerns e.g. small specimen size
optimisation based on FEM calculations, micro toughness modelling and qualification of small specimen test
technology towards accepted standards. In SPIRAL2, a deuteron beam of 5 mA and 40 MeV interacts with a
rotating carbon disk producing high energy neutrons (in the range between 1-40 MeV) via C(d,xn) reactions.
This facility, which will produce neutron-rich fission fragments for RNB physics studies, could be used for 3-4
months a year for material irradiation purposes. Estimations, taking into account this exposure time for a
fusion dedicated irradiation plug, lead to damage rates in the order of 1-2 dpa/y (in Fe) in a volume of ~10
cm3. Therefore the use of miniaturised specimens is essential in order to effectively utilize the available
irradiation volume in SPIRAL2. Sample package irradiation temperature would be in the range of 250 C to
1000 C. The irradiation level of 1 dpa/y with 14 MeV neutrons (average energy) may be interesting for microstructural and metallurgical investigations (e.g., mini-traction, small punch tests, etc.) and possibly for the
understanding of specimen size/geometric effects of critical material properties. Due to the small test cell
volume, sample in situ experiments are not foreseen. However sample packages would be, if required,
available each month after transfer in a special hot cell on site. The SPIRAL 2 project as well as the possible
implementation of the dedicated area for material irradiations is briefly presented including expected
irradiation characteristics.
Corresponding Author:
MOSNIER ALBAN
amosnier@cea.fr
CEA/DSM/DAPNIA, CEA-Saclay, 91191 Gif-sur-Yvette, France
343
- I - Materials Technology and Breeding Blankets.
P1C-I-10
SCIENTIFIC AND TECHNICAL FOUNDATIONS AND TECHNOLOGIES
OF REDUCTION OF MHD-RESISTANCE OF DUCTS WITH HEAVY
LIQUID METAL COOLANTS IN MAGNETIC FIELD OF BLANKET
AND DIVERTER OF TOKAMAK
PINAEV SERGEY, BEZNOSOV ALEXANDR (1) MURAVIEV EVGENI (2) ORLOV YURY (3)
(1) Nizhny Novgorod State Technical University, Minin st. 24, 603600, Nizhny Novgorod, Russia (2) Research and Design
Institute of Power Engineering, P.O. Box 788, 101000, Moscow, Russia (3) IPPE, Bondarenko sq. 1, 249020, Obninsk,
Kaluga Region, Russia
Developing of liquid metals as coolants of blanket and diverter of tokamaks can lead us to choose coolant with
higher safety standards than lithium. Heavy liquid metal coolants such as lead, gallium, eutectic lead-bismuth
and lead-lithium ensure higher safety because they don't burn in the air and don't react with water and steam
like alkaline metals. For cooling diverter channels possible choice is gallium, lead based coolants are
candidates for blankets. Electroinsulating coating formation on the inner surface of ducts is an efficacious
solution of high MHD-resistance problem. Heavy liquid metal coolants facilitate formation of oxygen based
electroinsulating coating and help maintain their stability. Structure of coating is oxygen-containing compound
of coolant, structural material and coolants impurities. Research of MHD flow of heavy liquid metal coolants
in a transverse magnetic field and methods of MHD-resistance reduction by electroinsulating coating due to
oxide layer formation on the inner surface of ducts are carried out in the department of “Nuclear and Thermal
Power Stations” of the Nizhny Novgorod State Technical University. Formation of oxygen electroinsulating
coating were executed by two main methods of oxygen delivery to pipes surface: injection of oxygencontaining gas mixture into the coolant flow; leading oxygen-containing gas mixture into expansion vessel
over the free surface of coolant, with further inflow of oxygen into the coolant and delivery with flow of
coolant to surfaces of structural material. Content of “free” oxygen into the coolant was controlled by galvanic
concentration cell. Conclusions of the most effective methods of electroinsulating coatings formation are based
on results of direct measure of MHD-resistance of different heavy liquid metal coolants. It has been proven
experimentally that value of MHD-resistance of heavy liquid metal flow in round steel ducts with formed
electroinsulating coatings in transverse magnetic field is between theoretical values for conductive walls and
fully nonconducting walls. Electroinsulating coatings created in heavy liquid metal coolants are able to
decrease the value of MHD-resistance by more than 5 – 10 times (depending on the coolant and the technology
of coatings formation).
Corresponding Author:
PINAEV SERGEY
pinaev@nntu.sci-nnov.ru
Nizhny Novgorod State Technical University, Minin st. 24, 603600, Nizhny Novgorod, Russia
344
- I - Materials Technology and Breeding Blankets.
P1C-I-26
EFFECT OF UNDERSIZED SOLUTE ATOMS ON MICROSTRUCTURE
CHANGE
RYAZANOV ALEXANDER, V.A.EGOROV-A, H.MATSUI-B
-a-Russian Research Center” Kurchatov Institute”,123182, Moscow, Russia, -b- Institute for Materials Research, Tohoku
University, Katahira 2-1-1, Aoba-ku, Sendai 980-8577, Japan
Vanadium-based alloys are considered as one of candidate structure materials for fusion reactors and so
understanding of physical mechanisms of an effect of solute atoms in these alloys on microstructure change is
very important for development of fusion material technology. Experimental investigations show that in
irradiated binary vanadium alloys V-A (A=Fe, Cr and Si) the number densities of self interstitial (SIA) loops
are found to be much higher that in pure vanadium. This indicates that solute atoms trap SIAs and enhance
dislocation loop nucleation. In the present paper, the di-atomic cluster nucleation model is extended to describe
the formation of SIA loops in irradiated binary vanadium alloys, including the effect of undersized solute
atoms on SIA loop nucleation and growth. In this model undersize solutes are considered to have strong
binding with SIAs and can act as the loop nucleation sizes. The suggested model takes into account also the
effect of solute segregation to loops and dislocation lines. The segregation of impurity atoms (undersized
solute atoms) at dislocation lines and SIA loops modifies the dislocation bias, the sign of the bias correction
being opposite to that of impurity misfit. Such bias modification affects the nucleation and growth SIA loops
too. The influence of these two factors on nucleation and growth dislocation loops in binary vanadium alloys
are presented here. It is shown that under irradiation the density of dislocation loops increases with increasing
concentration of undersized solute atoms and growth kinetics of SIA loops in these alloys is different too. The
predictions of the model and performed numerical calculations are compared with observed experimental data
on dislocation loop formation and growth under electron irradiation. It is found that the model is able to
describe the main features of the experimentally observed nucleation and growth of SIA loops in binary
vanadium alloys.
Corresponding Author:
RYAZANOV ALEXANDER
ryazanoff@comail.ru
Russian Research Centre"Kurchatov Institute",123182,Moscow,Kurchatov Sq.1,Russia
345
- I - Materials Technology and Breeding Blankets.
P1C-I-39
RADIATION INDUCED CONDUCTIVITY AND SURFACE
ELECTRICAL DEGRADATION OF PLASMA SPRAYED SPINEL FOR
NBI SYSTEMS
A. MOROÑO, AND E.R. HODGSON
Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain
Plasma sprayed spinel is being considered as a possible candidate material for the insulator rings of the ITER
neutral beam injector bushing. During ITER operation the insulating material will be subjected to a radiation
field due to the plasma and the NBI accelerator itself. The radiation will cause an increase in the electrical
conductivity (Radiation Induced Conductivity RIC) and possibly permanent radiation induced electrical
degradation both within the volume and/or on the surface. The electrical insulating rings in the bushing will
have one side facing high vacuum and the other exposed to a pressurized gas (SF6, N2, or dry air). The
electrical behaviour of the surface of the insulator under irradiation will depend strongly on the environment
(pressurized gas or high vacuum) surrounding the material. The paper describes experiments performed to
evaluate the RIC and surface electrical degradation of plasma sprayed spinel from LWK-PlasmaCeramic. The
experiments have been performed in the beam line of a 2 MeV Van de Graaff accelerator, where 10x10x0.7
mm2 LWK spinel samples were irradiated in high vacuum (10-6 bar) with 1.8 MeV electrons, at dose rates
from 7 to 70 Gy/s and temperatures between 50 and 350 C. The electron beam was perpendicular to the 10x10
mm2 faces. Platinum central and guard electrodes were sputtered on one of the 10x10 mm2 surfaces and a
single earth electrode on the other. The experimental set-up permitted an electric field of up to 1 MV/m to be
applied to the sample and to measure both volume and surface conductivity during and after irradiation . The
volume RIC at 70 Gy/s and 350 C for LWK spinel is 0.30x10-9 S/m. This value was not observed to change
for doses up to 10 MGy. However the material exhibits severe surface electrical degradation in vacuum when
heated up to temperatures between 150 and 400 C The threshold temperature for surface degradation depends
on the surface studied. For the surface perpendicular to the material growth direction the critical temperature
for significant electrical surface degradation in vacuum was found to be 400 C. In the case of the surface
parallel to the growth direction, the temperature for surface degradation was 150 C. As this represents the
surface which will be exposed to the vacuum, this material should not be used as an electrical insulator in
vacuum at temperatures higher than 150 C.
Corresponding Author:
A. MOROÑO
morono@ciemat.es
Euratom/CIEMAT Fusion Association, 28040 Madrid, Spain
346
- I - Materials Technology and Breeding Blankets.
P1C-I-43
BLANKET MANUFACTORING TECHNOLOGIES :
THERMOMECHANICAL TESTS ON HCLL BLANKET MOCKS UP
CACHON LIONEL, DIEPPOIS JEAN PAUL* TALAND RÉMI* LAFFONT GUY* POITEVIN YVES**
*CEA Centre de Cadarache, Bât. 204, DTN/STPA/LTCG, 13108 St PAUL lez Durance, FRANCE **CEA Centre de Saclay,
DM2S/SERMA/LCA , 91400 SACLAY, France
In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder
and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes
reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The
power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel
cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to
the complex geometry of these parts and the high level of pressure and temperature loading, thermomechanical phenomena expected in the “HCLL blanket concept” have motivated the present study. The aim of
this study, carried out in the frame of EFDA Workprogramme, is to validate the manufacturing technologies of
HCLL blanket module by testing small scale mock-ups under ITER TBM representative conditions.The first
step of this experimental program is the design & manufacturing of a relevant test section in the DIADEMO
facility, which was recently upgraded with an He cooling, taking the opportunity of synergies with the gascooled fission reactor R&D program. The second step will deal with the thermo-mechanical tests. This paper
focuses on the relevancy of the DIADEMO engineering design for HCLL thermomechanical tests . In
particular, the He loop is an isobaric device with an operating pressure of 80 bar. A blower allows an He flow
rate of 30 g/s, and the test section is fed by helium at a temperature between 300 and 400 C. After the test
section the He temperature can be higher than 500 C. To optimise the power balance of the loop, an heat
exchanger of 50 KW will be used. The inlet He temperature of the blower has to be lower than 60 C. So a
cooler was designed to evacuate 40 KW maximum, by mean of the glycol/water loop of DIADEMO. The
“DIADEMO HCLL” loop will be in operation at the end of 2004.
Corresponding Author:
CACHON LIONEL
cachon.lionel@cea.fr
CEA Centre de Cadarache, Bât. 204, DTN/STPA/LTCG, 13108 St PAUL lez Durance, FRANCE
347
- I - Materials Technology and Breeding Blankets.
P1C-I-58
HIGH ENERGY PROTON DEGRADATION IN KU1 QUARTZ GLASS
CONSTANTINESCU BOGDAN, -
We studied the 3 and 12.6 MeV proton - room temperature - irradiation-induced modifications on ultraviolet
transmission properties on KU1 quartz glass samples. We started with 3 MeV proton irradiation at room
temperature. Two 0,8 mm thick samples provided by CIEMAT, Madrid have been implanted using the
Bucharest HVEC Tandem accelerator at 8 „e 1013 and 1,5 „e1014 protons, respectively. The optical
transmission properties (absorption, transmission and reflectivity) in the UV region have been measured with a
Cary 4 VARIAN spectrophotometer. For the lower dose, absorption peaks at 215 nm and 240 nm, similar in
shape, but smaller in intensity to gamma irradiation case, can be observed. For higher dose, a supplementary
202 peak appeared, splitting the 215 nm peak. The 3 MeV protons produce considerable ionization, which is
the main cause of the energy loss at such low energies. The result of the overall ionization is the 215 nm band.
As concerning the number of induced defects, our doses, equivalent to 4 „e 1015 p/cm2 and 7,5 „e 1015 p/cm2
produced 4 „e 10-6 and 7,5 „e 10-6 dpa, repectively, in the 3 MeV protons range in quartz (80 ƒÝm),
equivalent to 5 „e 10-4 and 9 „e10-4 dpa for 1 cm. We continued our KU1 quartz glass studies using 12.6 MeV
proton irradiation. The superiority of 12.6 MeV irradiation as compared to 3 MeV is evident, due to the
relative homogeneity of the induced defects (vacancies produced) across the target depth. We irradiated a 0.8
mm thick KU1 sample in the same conditions as for 3 MeV irradiation, at a total dose of 2 x 1014 protons
(12.6 MeV energy). We observed the presence of 215 nm peak (due to both electron and nuclear collisions
stopping) and a big reduction of 240 nm peak (due only to nuclear collisions). We evaluated the dose rate of
our 12.6 MeV proton irradiations at 200 Gy/s and the total irradiation doses at 10 and 20 MGy. Comparing our
spectra (mainly the intensity of 215 nm peak) with the results for gamma and high energy electron irradiations,
we can conclude for the 12.6 MeV proton irradiation that the saturation effect in absorption is obtained after a
10 MGy dose, as compared with 4-5 MGy for gamma and with 11-12 MGy for electrons, suggesting the
ionization process is essential for defect absorption centers in all the cases.
Corresponding Author:
CONSTANTINESCU BOGDAN
bconst@ifin.nipne.ro
INSTITUTE OF ATOMIC PHYSICS, POB MG-6, Bucharest, Romania
348
- I - Materials Technology and Breeding Blankets.
P1C-I-85
EXPERIMENTAL STUDY OF LITHIUM MHD FLOW IN SLOTTED
CHANNEL FROM V-4TI-4CR ALLOY
LYUBLINSKI IGOR, V.A. EVTIKHIN, A.V. VERTKOV, N.I. EZHOV, V.M. SHCHERBAKOV
FSUE “Red Star”, 1a, Elektrolytniy Proezd, Moscow, 115230, Russia
MHD pressure drop in flowing liquid metal for a tokamak with high magnetic fields is a key concern regarding
the development of lithium self-cooled test module for ITER and lithium breeding blanket for DEMO-type
projects. MHD losses on liquid metal pumping can be most efficiently reduced by applying of an electrically
insulating coating to the inner surface of the channels. The choice of materials and coating applying
technology requires an experimental procedure for the assessment of coating characteristics in conditions close
to reactor conditions. The developed method of estimation of electrically insulating coating properties on the
V-4Ti-4Cr channel internal surface is based on the pressure drop measurement in liquid lithium forced
circulation system with MHD test section. The tests were conducted on channels from V-4Ti-4Cr alloy with
insulating coating based on AlN and without coating. The dimensions of the test section were 6×20×320 mm.
Data were taken at lithium flow velocity up to 5 m/s, temperature up to 500oC and a uniform transverse
magnetic field up to 1.6 T. Measured hydraulic resistance ? depending on the MHD interaction parameter
Ha2/Re have shown five times reduction for coated wall in comparison with conducting wall channel.
Methods of improving the electrically insulating coating characteristics on the vanadium alloys are considered.
Corresponding Author:
LYUBLINSKI IGOR
lyublinski@mtu-net.ru
FSUE “Red Star”, 1a, Elektrolytniy Proezd, Moscow, 115230, Russia
349
- I - Materials Technology and Breeding Blankets.
P1C-I-88
A NEUTRONIC INVESTIGATION OF HE-COOLED LI-BREEDER
BLANKETS FOR FUSION POWER REACTOR
KIM, YONGHEE, HONG, BONG GUEN
150 Deokjin-dong, Yuseong-gu, Daejeon 305-353, Republic of Korea
In Korea, a liquid metal blanket is being studied as an option of the ITER test blanket. The R&D efforts are
tuned to He-cooled and Li-breeder blankets after assessment of various liquid metal blanket concepts. Major
technical rationale for the selection is in that the concept is virtually free from the tritium (T) permeation
problem and the MHD-related issues. This paper is concerned with neutronic investigation of the liquid metal
blanket concepts. The He-cooled blanket is adopting a multi-layered design concept in the radial direction.
Both the first wall and the breeding zone are cooled by the He coolant. In the design, a thin Li layer is placed
between the first wall and the coolant in order to prevent tritium permeation into the coolant channel from the
plasma zone. For an efficient T breeding, a static neutron multiplier is also introduced into the blanket. In order
to minimize the neutron leakage to the vacuum vessel (VV), a neutron reflector and a neutron absorber are
placed after the breeding zone. Finally, a gamma shield is put between the absorber and VV. For T recovery,
the Li breeder is circulated very slowly such that the MHD pressure drop might not be an issue. Various
optimization studies have been performed with a neutron transport code in a one-dimensional cylindrical
geometry mainly from the neutronic point of view. As the performance measure of the blanket, three design
parameters were considered: tritium breeding ratio, Li-6 enrichment (or Li volume), and the energy
multiplication in blanket. Three neutron multiplier options (Be, PbO, and W) were evaluated in terms of the
three performance measures. In a fusion reactor, it is crucial to maintain a self-sustaining tritium cycle.
Meanwhile, energy production of the (n,T) reaction of Li-6 is quite significant in a T-self-sufficient cycle (
about 4.8 MeV per Li-6 (n,T) reaction). To investigate the impact of neutron spectrum on the blanket
performance, a graphite moderator was assessed in this paper. Also, the blanket performance was also
evaluated in terms of the Li-6 enrichment. In addition, conversion of Li-7 to Li-6 was investigated.
Corresponding Author:
KIM, YONGHEE
yhkim@kaeri.re.kr
150 Deokjin-dong, Yuseong-gu, Daejeon 305-353, Republic of Korea
350
- I - Materials Technology and Breeding Blankets.
P1C-I-96
MICROSTRUCTURAL CHARACTERISATION OF EUROFER-ODS
RAFM STEEL IN THE NORMALIZED AND TEMPERED CONDITION
AND AFTER THERMAL AGING IN SIMULATED FUSION
CONDITIONS
PAÚL, ANTONIO, O. M. MONTES (1) E. ALVES (2) L. C. ALVES (2) R. LINDAU (3) J. A. ODRIOZOLA (1)
(1)Instituto de Ciencia de Materiales de Sevilla. Avda. Américo Vespuccio s/n, 41092 Sevilla, Spain (2) ITN, Estrada
Nacional 10, Sacavém Portugal (3) EFDA, Garching, Germany
ODS RAFM steels are promising candidates to be used as structural materials in fusion reactors, mainly due to
their creep and swelling resistance. In this work we present the results of our research on the microstructure of
EUROFER based ODS using different characterization techniques. Preliminary results with optical
microscopy on the as-received material indicate that the microstructure is ferritic. In addition, ion microprobe
studies reveal a homogeneous distribution of yttrium particles in the ferritic microstructure. The austenitisation
temperature will be determined by in-situ high temperature XRD in order to find the most adequate
normalization and tempering treatments to obtain a fully ODS martensitic microstructure. Samples in the
normalized and tempered condition will be characterized by means of XRD, optical microscopy, SEM and
TEM so that a complete characterisation of the microstructure will be obtained. Special attention will be paid
to the yttrium oxide dispersion and the presence of precipitates. During operation in nuclear fusion plants the
structural materials will be exposed to high temperatures. Experiments of thermal aging at 700 C in He/H gas
mixture up to about 5000 hours are being performed. The microstructure after thermal aging will be compared
to that of the original material. Relevant results of this research will be presented in the conference.
Corresponding Author:
PAÚL, ANTONIO
momo@icmse.csic.es
Instituto de Ciencia de Materiales de Sevilla. Avda. Américo Vespuccio s/n, 41092 Sevilla, Spain
351
- I - Materials Technology and Breeding Blankets.
P1C-I-102
NON-DESTRUCTIVE ANALYSIS OF MINIATURIZED FUSION
MATERIALS SAMPLES AND IRRADIATION CAPSULES BY X RAY
MICRO-TOMOGRAPHY
TISEANU ION, TEDDY CRACIUNESCU BOGDAN N. MANDACHE
National Institute for Laser, Plasma and Radiation Physics Plasma Physics and Nuclear Fusion Laboratory
Recently, at the Association EURATOM-MECT (Romania) a laboratory for X-ray microtomography was
established with European Community support. Its research is focused on NDT inspection of miniaturized
samples of fusion materials and irradiation capsules for IFMIF environment conditions. Computer-aided
tomography (CAT) systems are configured to take many views (radiographies) of the object in order to build a
3-D model of its internal structure. X-ray tomography as an NDT tool for fusion material samples can provide
information on: density variations, micro-cracks development by mechanical/thermal cycling, permeability of
porous materials, components microstructure integrity, 3-D accurate geometrical measurements. Our
tomographic facility consists of an open type microfocus X-ray source, a five axis micrometric
translation/rotation manipulator and optionally large area, high resolution image intensifier or amorphous
silicon flat panel as X-ray detection system. This setup permits high resolution cone-beam tomography of
miniaturized samples as well as an innovative oblique view inspection of the flat samples or components as
irradiation capsules, IFMIF Li-target backplate etc. 3-D tomographic reconstructions are obtained by a
proprietary computer code based on a modified Feldkamp algorithm. The reconstruction software also
incorporates efficient techniques for beam hardening reduction and ring artifacts elimination. By numerous
experiments it was established that our system can be used for a large range of samples with regards to size,
material and complexity. For the individual miniaturized samples the microtomography analysis is guaranteed
for feature recognition down to few microns. A space resolution of tens of microns for irradiation capsules of
around 100 mm characteristics dimension is currently obtained. The microtomography facility is available for
EFDA Technology Programme. The future activities will be focused on CAT structural integrity assessment of
instrumented capsules and rigs and the development of real-time micro-radiography of miniaturized samples
under mechanical/thermall stress. In addition to the transmission tomography studies one presents a conceptual
design of an emission tomographic system for already irradiated miniaturized samples and capsules.
Numerical simulations validated by experimental tests show that the design parameters for space resolution
and isotope selectivity are well within reach.
Corresponding Author:
TISEANU ION
tiseanu@alpha2.infim.ro
National Institute for Laser, Plasma and Radiation Physics, Plasma Physics and Nuclear Fusion Laboratory,
Atomistilor Str. No 111, P.O. Box MG-36, R-76900 Bucharest, Magurele, Romania
352
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P1C-I-108
INNER STRUCTURES OF COMPRESSED PEBBLE BEDS
DETERMINED BY X-RAY TOMOGRAPHY
REIMANN JOERG, 1)R. A. PIERITZ, 2)M. DI MICHIEL, C. FERRERO
1) Applied Research Solutions, 15 place du Charmeyran, F-38700 La Tronche, France 2) European Synchrotron Radiation
Facility,B.P. 220,F-38043 Grenoble CEDEX, France
In the Helium-cooled Pebble Bed (HCPB) blanket, beryllium in form of pebbles is planned to be used as a
neutron multiplier. During operation, thermal stresses will result in a compression of these beds which
influences significantly the pebble bed thermal conductivity. For the blanket design, the accurate knowledge of
the dependence of this thermal conductivity on the compression state is of large importance. Currently, this
dependence is measured in uniaxial compression tests considering the pebble bed as a “black box”. For the
extrapolation of data and the improvement of currently available heat transfer correlations the knowledge of
the number of pebble contacts and corresponding contact zones within the bed is of great interest. Experiments
were performed first in the Forschungszentrum Karlsruhe where cylindrical pebble beds were pre-compressed
to different strain levels in uniaxial compression tests. For higher measurement accuracy, spherical 3.5
aluminium pebbles were used instead of the (for fusion applications) usual1 mm beryllium pebbles. In the
European Synchrotron Radiation Facility (ESRF) Grenoble, a special microtomography experimental setup
was then used allowing the computer aided reconstruction of 3-D images of the attenuation coefficient of the
X-ray synchrotron radiation beam within the pebble beds. By post-processing the data, very useful information
was obtained on both radial and axial void fraction distributions in the samples as well as the detailed
information on pebble contact numbers and contact zones. In the paper, the microtomographic technique as
well as the first results of the analyses are presented and critically discussed in special regard to future
investigations.
Corresponding Author:
REIMANN JOERG
joerg.reimann@iket.fzk.de
Forschungszentrum Karlsruhe, Institut für Kern- und Energietechnik, P.O. Box 3640, D-76021 Karlsruhe,
Germany
353
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P1C-I-109
THERMAL CREEP OF BERYLLIUM PEBBLE BEDS
HARSCH HEINRICH, JOERG REIMANN
Forschungszentrum Karlsruhe, Institut für Kern- und Energietechnik, P.O. Box 3640, D-76021 Karlsruhe, Germany
Present ceramic breeder blanket designs are based on ceramic breeder and beryllium pebble beds. During
operation, thermal stresses arise from different thermal expansions of the pebble beds and structural materials,
and pebble bed swelling due to irradiation. Thermal creep of pebble beds will partly release the build-up of
stresses, might improve heat transfer due to increased contact areas between the pebbles and might compensate
a further stress build-up due to irradiation induced swelling. Therefore, the knowledge of thermal creep is of
prime importance for both types of pebble beds. In the past, thermal creep investigations were restricted to
different granular ceramic breeder materials In this paper, first results for beryllium pebble beds consisting of
1mm NGK pebbles are presented. These experiments were performed in the uniaxial test facility HECOP II in
a temperature range between 450 and 650 C and uniaxial stresses up to 3.6MPa. Thermal creep strain was
described by a correlation of the type ecr = A exp(B/T) sp tn. Compared to ceramic breeder materials, both the
coefficient A and exponents p and n differ. However, it is interesting to note that for relatively short creep
periods and blanket relevant temperatures, the creep strain is very similar for Li4SiO4 and beryllium pebble
beds. With these new results, for the first time a complete set of data exists required for the description of the
thermomechanical interaction of solid breeder and beryllium pebble beds with the structural material of
blanket elements.
Corresponding Author:
HARSCH HEINRICH
heinrich@versuchstechnik.de
Goraieb Versuchstechnik, In der Tasch 4a, D-76227 Karlsruhe, Germany
354
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P1C-I-110
THERMAL CREEP BEHAVIOR OF THE EUROFER97 RAFM STEEL
AND TWO EUROPEAN ODS-EUROFER97 STEELS
NADINE BALUC, NOBUYASU NITA (1) GANG YU (2)
(1) Matsui Lab., Institute for Material Research Tohoku University Katahira 2-1-1, Aoba-ku Sendai, Japan 980-8577 (2)
Fusion Technology-Materials, CRPP - EPFL , AssociationEURATOM-Confederation Suisse, 5232 Villigen PSI, Switzerland
The reduced activation ferritic/martensitic (RAFM) steel EUROFER97 and the oxide dispersion strengthened
(ODS) steel ODS-EUROFER97, with the EUROFER97 as matrix material and 0.3 wt.% Y2O3 particles as
reinforcement material, are foreseen to be used as structural materials in fusion power reactor at temperatures
up to about 550 C and 650 C, respectively. Their creep behavior is one of the key issues for their future
application. Thermal creep tests have been conducted on the EUROFER97 and two kinds of ODSEUROFER97, which were manufactured using slightly different powder metallurgy procedures, at the Centre
of Research in Plasma Physics (Switzerland) and at the CEA-Grenoble (France), respectively. Thermal creep
experiments were conducted under constant stress at temperatures between 450ºC and 750ºC, in an argon flow,
up to rupture. They were complemented with post-testing microscopic observations of the specimen and
rupture surfaces. It was found that the ODS-EUROFER97 exhibits significantly higher creep strength than the
EUROFER97, and could clearly be used 100 degrees above the EUROFER97. Creep exponents have been
determined. A creep exponent of about 4 was found for the ODS-EUROFER, which is characteristic of a climb
dislocation mechanism at obstacles, i.e. the Y2O3 particles. A creep exponent of about 14 was found the
EUROFER97, which indicates that the stress sensitivity of the strain rate is much less for the ODSEUROFER97 than for the EUROFER97, which is beneficial for its future use.
Corresponding Author:
NADINE BALUC
nadine.baluc@psi.ch
Fusion Technology-Materials, CRPP - EPFL , AssociationEURATOM-Confederation Suisse, 5232 Villigen PSI,
Switzerland
355
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P1C-I-122
SEGREGATED VOID SWELLING IN A HETEROGENEOUS
MATERIAL: IMPLICATIONS FOR FUSION MATERIALS
SERGEI DUDAREV (1), ALEXEI SEMENOV(2) AND CHUNG WOO(2)
(2)Department of Mechanical Engineering, Polytechnic University of Hong-Kong, Hung Hom, Kowloon, Hong-Kong
Nucleation and growth of voids and gas bubbles in materials under irradiation is known to represent one of the
key factors limiting the lifetime of structural materials under irradiation. While in many cases nucleation and
growth of voids is uniform and spatially homogeneous, the occurrence of large-scale spatially heterogeneous
distributions of voids has long defied a natural theoretical explanation. This now impedes the understanding of
the behaviour of technological alloys and steels irradiated by fusion neutrons, since these materials have
spatially heterogeneous microstructure. We report new theoretical investigations of the nucleation and growth
of voids in a neutron irradiated copper with spatially heterogeneous dislocation microstructure. We aim to
understand the origin of spatially segregated void swelling observed in several materials in the limit of
relatively low irradiation dose. Experimental observations show that voids form predominantly in the regions
of low density of dislocations, in a seemingly apparent contradiction with the reaction-diffusion “Standard
Rate Theory” model, in which growth of voids is driven by the preferential absorption of interstitial atoms by
dislocations. We find that, due to the high sensitivity of the void nucleation rate to the local supersaturation of
vacancies, voids nucleate and grow almost exclusively in the regions where the density of dislocations is low.
The model shows that the relatively high void swelling rates observed experimentally in the regions of low
dislocation density can be naturally explained by taking continuous nucleation of voids into account. There are
two aspects of this work that have implications for the design of new materials for fusion technology. On the
one hand, the new results show that the diffusion model of transport of interstitial defects provides an adequate
framework for the treatment of swelling of a spatially heterogeneous material. On the other hand, the new
approach suggests ways of treating microstructural evolution of complex composite materials e.g. multiphase
steels that have so far been only investigated using spatially homogeneous mean-field models.
Corresponding Author:
SERGEI DUDAREV (1)
sergei.dudarev@ukaea.org.uk
(1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Oxfordshire OX14 3DB, UK
356
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P1C-I-128
THERMOCHEMISTRY OF LI-TITANATES CERAMICS IN REDUCING
ENVIRONMENTS
MANCINI MARIA RITA, V. CONTINI(1), K. TSUSHIYA(2), R. GIORGI(1), F. PIERDOMINICI(1), E.
SALERNITANO(1), T. HOSHINO (2), H. KAWAMURA(3) AND S. CASADIO(1),
(1) ENEA, CR Casaccia, via Anguillarese, 301, 00060 S.M. di Galeria, Roma, Italy (2) JAERI, Oarai-machi, HigashiIbaraki-gun, Ibaraki-ken, 311-1394, Japan (3) JAERI, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193, Japan
Temperatures close to 1273K are envisaged to occur in the core of the tritium breeding Li2TiO3 pebble bed
conceived for the European and Japanese Blanket designs of fusion reactors, while purged by He+0.1%H2 gas.
On the mean time a lithium depletion due to Li transmutation is expected to reach a burn-up of 18%. To
simulate this “hot spot” situation a set of accelerated tests were performed on ad-hoc prepared ceramic pellets
in which Li/Ti atomic ratio was decreased by doping Li2TiO3 with TiO2, leading to the formation of
Li4Ti5O12 spinel phase. These were exposed to Ar + 3% H2 purge gas at 1173 and 1273K for 15 hours by
recording their weight loss in a Pt crucible of a Thermo-Gravimetry (TG) and Differential Thermal Analysis
(DTA) apparatus. The Li-titanate reduction amount and its TiO2 doping degree were found to be
approximately related, and the higher the Li-depletion (i.e. the increasing amount of spinel phase) the higher
was the reactivity. The reaction resulted to be complex, since lithium loss by evaporation under reductionannealing was evidenced. Therefore, some of the reduced specimens were completely re-oxidized in flowing
air while undergoing a fast ramp annealing up to 773K in the thermo-balance. Both fine powders and sintered
pellets of near stoichiometric spinel were tested as described above. Reduction annealing at 1173K was found
to evolve without significant modification of the cubic phase. On the contrary, the 1273K run was found to
induce a significant structural modification of both the powder and pellets specimens. They were found to
decompose into an orthorhombic Li0.14TiO2 (strongly Li-depleted, well detected by XRD) and a Li-rich
phase which was forced to move to the surface of the specimen where an intense lithium signal could be well
detected by an XPS. The reported results reveal: i) that the H2 reaction with pure Li2TiO3 at 1173K is
negligible; ii) that this reaction is enhanced as the “initial” Li-depletion (or TiO2-doping) increases in the
specimens; iii) that Li loss by evaporation also increases on the mean time inducing a complex synergy with
the O-vacancy generation. There is an extreme phase evolution of the spinel phase at 1273K. Discussion is
reported about how this phenomenon influence the Li-depleted Li-meta-titanate pebbles, that being of interest
to relate pebbles degradation with Li-burnup due to the H2 added to the He purge gas at the highest envisaged
temperatures of the HCPB Blankets.
Corresponding Author:
MANCINI MARIA RITA
rita.mancini@casaccia.enea.it
ENEA, CR Casaccia, via Anguillarese, 301, 00060 S.M. di Galeria, Roma, Italy
357
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P1C-I-129
MOLECULAR DYNAMICS SIMULATIONS OF DEFECT PRODUCTION
DURING IRRADIATION IN SILICA GLASS
MOTA, FERNANDO, M.-J.CATURLA(2),J.M.PERLADO(1), E.DOMINGUEZ(1), A.KUBOTA(3)
(1) Instituto de Fusión Nuclear, Universidad Politecnica , Madrid, Spain. (2)Universidad de Alicante, Dep.Física Aplicada,
Alicante, Spain (3)Lawrence Livermore National Laboratory, Livermore CA, USA.
Irradiation of materials such as fused silica can change not only its mechanical properties but also its optical
properties through the creation of colour centers. Identifying and understanding the type of defects created
during irradiation is a complicated experimental task that could benefit from some insight obtained from
atomistic simulations. In this paper we present molecular dynamic simulations of defect production in silica
glass. Displacement cascades with Primary Knock-on Atom (PKAs) energies between 1 and 10 keV have been
simulated. We have studied the production of Oxygen deficient centers (ODCs) and other type of defects
possible , as well as, its mechanism for production and recombination at short time scales. We compare the
number of defects produced with experimental data whenever available, and we extract consequences
regarding its optical and mechanical properties under irradiation.
Corresponding Author:
MOTA, FERNANDO
mota@denim.upm.es
Universidad Politécnica de Madrid, Escuela Tecnica Superior de Ingenieros Industriales, Instituto de Fusión
Nuclear C/José Gutierrez Abascal 2 CP:28006 Madrid (SPAIN)
358
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P1C-I-130
KINETICS OF LI DEPLETED LI2TIO3 REACTION WITH H2 ADDED
TO AR PURGE GAS
CONTINI VITTORIA, C.ALVANI(1), K.TSUSHIYA(2), F. PIERDOMINICI(1), H. KAWAMURA(3) AND S.
CASADIO(1)
(1) ENEA, CR Casaccia, via Anguillarese, 301, 00060 S.M. di Galeria, Roma, Italy (2) JAERI, Oarai-machi, HigashiIbaraki-gun, Ibaraki-ken, 311-1394, Japan (3) JAERI, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193, Japan
Actual fusion DEMO reactor blanket designs are considering lithium-titanate (Li-Ti) pebble bed in which the
peak temperature may be critical. Li-Ti-pebbles are known to be reduced when exposed to He + H2 (0.1%)
purge (R-gas) by the reaction scheme (1), where w is the consumed H2 (mol) = generated H2O (mol) per Ti gatom and the O-vacancies (atom) generated in the pebbles [black box] H2 + Li-Ti-pebbles = wH2O+ (1-w)H2
+ [black box] (1) The product [black box] is unknown as well as the extension of (1) as a function of other
parameters. For example Li-depletion may be of concern, since a Li burn-up ~18% is predicted for European
HCPB blanket. The kinetics of (1) was investigated on pebbles and pellets with Li/Ti atom ratio was decreased
by doping lithium meta-titanate with TiO2 generating Li4Ti5O12 spinel phase. The reduction (1) was
performed in R-gas and Ar + 3%H2 at 1173 K. Soon after these TPR runs the specimens were re-oxidized by
ramp-heating them in Ar + H2O (1.5%), reversing process (1) which anneals the O-vacancies previously
generated (TPO runs). This method was tested versus the standard TPO using the oxidizing He + O2 (0.1%)
and the Ti(III) to Ti(IV) valence conversion was found to be equivalent in the two cases. The rate of (1) was
found improved by decreasing the Li/Ti ratio. A diffusion controlled reaction mechanism was found to fit the
data better than gas-solid interface rate controlling step. TPR and the TPO data were not found in agreement
suggesting that a part of the H2 consumed during TPR induced some Li-loss. XRD analysis of specimens
reduction-annealed for 17 h showed no significant changes in the structures, although SEM analysis showed
the nucleation of a phase at the grain surface (not XRD -detectable) of the specimens containing spinel
phase.fraction (=5ƒ) > 1%. Hence Li-depleted Li-Ti-pebbles were found determining the rate of (1) and the
black box in the scheme (1) was found well described by [black box] = [(1-ƒ)Li2TiO3-x + ƒLi4/5TiO4-y.]
with w = (1-ƒ)x+ƒy (2) where x and y refer to O-vacancy mol per Ti g-atom in Li-metatitanate and spinel
phase respectively. After 17 hours at 1173 K the pure spinel (ƒ= 0; w =x) and of pure Li-metatitanate (ƒ = 1; w
= y) showed the values (x = 0.2 %, y = 4 %) in R-gas and (x = 2 %, y = 10 %) in Ar+3% H2 gas., and the
results for the intermediate Li-depleted specimens could be fitted with (2) within the experimental error.
Corresponding Author:
CONTINI VITTORIA
vittoria.contini@mail.casaccia.enea.it
ENEA, CR Casaccia, Via Anguillarese 301, 00060 S. Maria di Galeria, Roma, Italy
359
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P1C-I-134
VITAMIN-J/COVA/EFF-3 CROSS-SECTION COVARIANCE MATRIX
LIBRARY AND ITS USE TO ANALYSE BENCHMARK EXPERIMENTS
IN SINBAD DATABASE
KODELI IVAN-ALEXANDER,
The paper will present the preparation, testing and use of the new cross-section covariance matrix library ZZVITAMIN-J/COVA/EFF3 intended for use in the sensitivity and uncertainty analysis. The library includes: Cross-section covariance data from the EFF-3 evaluation for the following 5 materials: Be-9, Si-28, Fe-56, Ni58 and Ni-60. - FORTRAN program ANGELO2 for the extrapolation/interpolation of the covariance matrices
to another energy group structure. - FORTRAN program LAMBDA to verify the mathematical properties of
the covariance data. The covariance matrices were processed by the NJOY97.115 code to the energy group
structure given in ENDF files and stored in the highly compressed BOXER format. The matrices were checked
for negative eigenvalues using the LAMBDA code, and all the File-33 covariance matrices were found
positive. The ANGELO2 code is then used to interpolate the covariance data linearly in lethargy to the group
structure, as specified by the user. The motivation for this work was the believe that in case of EFF-3
covariance data the use of the covariance matrix library and the ANGELO2 code could be more convenient for
the users than the processing by the NJOY code. In particular ANGELO-2 code is easy to use and fast, and
permits to avoid some difficulties of the NJOY processing. On the other hand no flux or cross-section
weighting is used in the interpolation process. The library was used as part of the SUSD3D based sensitivity
and uncertainty computational package to analyse several fusion experiments included in the SINBAD
collection of benchmark data. The uncertainties based on the cross-section covariance data were compared
with the observed C/E values, as well as with those based on the covariance information from other
evaluations like ENDF/B-VI. The collapsing procedure used in the ANGELO-2 code was also compared with
the one used in the NJOY system. Among the benchmarks available in the SINBAD package (altogether 34
covering Reactor Shielding, 22 Fusion Neutronics Shielding and 8 Accelerator Shielding) the following were
analysed: - FNG-ITER Blanket Bulk Shield - FNG-ITER Neutron Streaming - FNG Silicon Carbide In
addition the library is being used in the pre-analysis of the future TBR benchmark in preparation at the FNG
Frascati. The library was prepared in the scope of the European Fusion Technology Programme and is now
available from the NEA Data Bank as the package NEA-1264/05.
Corresponding Author:
KODELI IVAN-ALEXANDER
ivo.kodeli@oecd.org
OECD NEA Data Bank, 12 Bd des Iles, F-92130 Issy-les-Moulineaux, France
360
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P1C-I-140
IN-SITU FORMATION AND CHEMICAL STABILITY OF ER2O3
COATING ON V-4CR-4TI IN LIQUID LITHIUM
YAO ZHENYU, SUZUKI AKIHIRO (1) MUROGA TAKEO (2)
(1) Nuclear Engineering Research Laboratory, University of Tokyo, Shirakata-Shirane 2-22, Tokai-mura, Naka-gun, Ibaragi
319-1188, Japan (2) National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, Japan
A self-cooled Li/V-alloy blanket is an attractive concept for future fusion reactors. One of the critical issues
for the blanket is the magneto-hydrodynamic (MHD) pressure drop. The electrical insulating coatings on the
inner wall of the blanket components for mitigating the pressure drop are under development. The in-situ
formation in liquid lithium is a quite attractive technology because it will make the coating on complex
surfaces possible and has potentiality to heal the cracks of coating without disassembling. The Er2O3 ceramic
shows high thermodynamic stability, good compatibility with liquid lithium, good stability in air and high
electrical insulating property, and thus is regarded as a candidate material. In the previous study, the authors
showed in-situ chemical formation of Er2O3 layer by exposing V-4Cr-4Ti in liquid Li doped with Er at high
temperature. In this paper, the characteristics and the long-term stability of the coating were investigated. The
V-4Cr-4Ti were oxidized at 973K, annealed at 973K, and finally exposed to liquid Li doped with Er at 873 K
and 973K. The oxygen was charged at surface by pre-oxidation and homogenized to the limited depth of 150
microns by subsequent annealing. During oxidation, nitrogen was not introduced into the specimens. At 873K,
the coating was stable up to 750h of exposure in liquid lithium. The thickness of Er2O3 layer saturates at about
0.1 micron with increasing exposure time. No cracking in the coating occurred during the cooling. The
resistivity of coated V-4Cr-4Ti increased about 12 orders of magnitude comparing with bare substrate and
exceeded 10e+13 ohm-cm. The sample exposed at 873K was subsequently exposed in liquid lithium doped
with erbium at 973K, the thickness of Er2O3 layer increased to 0.4 micron. By isothermal exposure at 973K,
the thickness of Er2O3 formed was 1.1 micron. The formation and growth of the coating were investigated as
a function of temperature and time of the exposure and the Er doping level to Li.
Corresponding Author:
YAO ZHENYU
yao@nifs.ac.jp
Department of Fusion Science, School of Physical Science, The Graduate University for Advanced Studies, 322-6
Oroshi, Toki, Gifu 509-5292, Japan
361
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P1C-I-141
PHYSICO-CHEMICAL PROPERTIES OF AND HYDROGEN ISOTOPE
BEHAVIORS IN LITHIUM-TIN ALLOY AS A LIQUID BREEDER FOR
FUSION REACTOR
KANG YI, TERAI TAKAYUKI
Department of Quantum Engineering & Systems Science School of Engineering, The University of Tokyo 2-11-16 Yayoi,
Bunkyo-ku, Tokyo 113-8656, Japan
Lithium-tin alloy is a potential material for tritium breeder and coolant in the liquid wall concepts of fusion
reactor. However, many properties of lithium-tin have not been measured, which are very important to
evaluate the feasibility of its application to fusion reactor systems. Some measurements were carried out to
complete the database of this material on the physico-chemical and tritium properties, such as the density and
the vapor pressure of the liquid alloy, the tritium diffusivity and the hydrogen solubility in the alloy. A
research on the tritium diffusivity, furthermore deducing the possible solubility range, has been investigated by
the in-reactor tritium release experiments using the fast neutron source reactor “YAYOI” of the University of
Tokyo. The experiments were performed at the reactor power of 500 W and 2 kW, from 723K to 873K and
with H2 partial pressure varying from 1.1 to 101 k Pa in He purge gas. The order of the diffusion coefficient is
10-9 m2/s at 873K. The density of the Li20Sn80 between 673K and 873K was obtained by using the
Archimedean principle. It can be expressed by the equation: Density[g/cm3] = 6.380 – 4.745×10-4 T [K] (673
– 873K) The hydrogen solubility measurement was conducted at 873 K. the method used in this studies is
based on pressure increase measurements in a known volume during desorption from the liquid sample. The
preliminary results showed it has higher solubility than Li17Pb83 alloy and near the solubility of hydrogen in
Sn. The vapor pressure of the Li20Sn80 alloy is carried out by Knudsen Cell mass spectrometer in the
temperature range of 773 to 1073 K. The main vapor species was Li, and Sn vapor was very low.
Corresponding Author:
KANG YI
kangyi@starling.q.t.u-tokyo.ac.jp
Department of Quantum Engineering and Systems Science, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku,
Tokyo 113-8656, Japan
362
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P1C-I-143
INTEGRAL EXPERIMENT ON BERYLLIUM WITH D-T NEUTRONS
FOR VERIFICATION OF TRITIUM BREEDING
VERZILOV YURY, OCHIAI KENTARO, KLIX AXEL, SATO SATOSHI, WADA MASAYUKI AND NISHITANI TAKEO
Numerous clean experiments on beryllium have been performed with the main objective of investigating the
neutron multiplication issues. Significantly less attention was paid to tritium breeding due to the former design
concept. At the moment, it is realized that the beryllium neutron multiplication power can be suitably predicted
in design calculations of breeder blankets. However, observed discrepancies between calculations and
experiments for the spectral flux distribution can affect the tritium breeding ratio. This is especially important
for the present design of the solid breeder blanket of the DEMO reactor, since the resulting neutron flux
spectra for tritium breeding is mostly formed by beryllium. In order to fulfill the necessity for experimental
verification of the essential parts of the neutron spectra vital from the tritium breeding point of view, clean
integral experiment was performed. An experimental assembly was constructed from beryllium blocks (S-200F, Brush Wellman Inc., USA) and shaped as a pseudo-cylindrical slab with an area-equivalent diameter of 628
mm and a thickness of 355 mm. The reaction rates such as 31P(n,g)32P, 6Li(n,a)3H, 32S(n,p)32P and 7Li(n,
n`a)3H were measured using activation detectors at various depths inside the assembly. The reactions
31P(n,g)32P and 32S(n,p)32P are characterized by the same energy response, as the tritium production
reactions on the lithium isotopes. In addition, 92Nb(n,2n)92mNb reaction rate was measured to verify the
neutron source distribution. In order to minimize the perturbation effect of the thermal neutron flux in the
beryllium assembly by the activation detectors, thin Li2CO3 detectors with a low concentration of 6Li were
used. The experimental analyses were performed using the Monte Carlo code MCNP-4C with the
FENDL/MC-2.0 and JENDL-3 nuclear data libraries. From the obtained results, the nuclear data files were
confirmed to be fairly reliable with respect to the prediction of the tritium breeding. Based on the obtained
results and previous benchmark experiments, it is possible to conclude that the beryllium data evaluations are
acceptable for neutron fusion applications.
Corresponding Author:
VERZILOV YURY
verzilov@fnshp.tokai.jaeri.go.jp
JAERI, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 JAPAN
363
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P1C-I-147
CREEP STRENGTH OF REDUCED ACTIVATION
FERRITIC/MARTENSITIC STEEL EUROFER'97
P. FERNÁNDEZ, A.M. LANCHA 1), J. LAPEÑA 1), R. LINDAU2), M. RIETH2) AND M. SCHIRRA2)
1)CIEMAT, Avda. Complutense nº 22, 28040 Madrid (Spain) 2)Forschungszentrum Karlsruhe, Institut für Materialforschung
I, P.O. Box 3640, 76021 Karlsruhe (Germany)
Since several years a new European reference structural material for fusion power systems, denominated
Eurofer'97 steel, is being intensively investigated. The continuos development and qualification of reduced
activation ferritic/martensitic steels for fusion applications require an exhaustive understanding of their
microstructure and mechanical properties. Of special relevance is the behaviour of these materials under long
term loading conditions at the high temperatures of fusion reactor operation. Therefore, the creep properties of
these steels and their microstructural evolution during creep must be studied, because the maximum operating
temperature of the fusion power plants will be determined, among others, by the creep characteristics. The aim
of this work is to evaluate the creep rupture strength properties of the Eurofer'97 steel in different product
forms (plate and bar) in the as-received condition: normalized at 980 C/27' plus tempered at 760 C/90'/aircooled for the plate, and normalized at 980 C/110' plus tempered at 760 C/220'/air-cooled for the bar. Creep
tests are being carried out in the temperature range from 450 C to 650 C at different loads, from 370 MPa to 50
MPa. At present, some creep tests (long testing times) of Eurofer'97 are still running. In addition, the creep
behaviour of the Eurofer'97 steel is being compared with the creep properties of the reduced activation
ferritic/martensitic steel F-82H mod. previously studied. Several different assessments (Norton law,
Monkman-Grant relation and Larson-Miller parameter) are being performed to evaluate the creep rupture
properties and to estimate the long-term creep rupture strength behaviour of the Eurofer'97 steel. No
significant differences in the creep rupture properties have been found between the different product forms
investigated (plate and bar). The Eurofer'97 has shown adequate creep rupture strength levels at short creep
rupture tests, similar to those of the reduced activation ferritic/martensitic steel F-82H mod. However, for long
testing times (> 9000 h) the results available up to now at 500 C and 550 C seem to indicate that a different
degradation mechanism of the creep properties is taking place in the Eurofer'97 steel.
Corresponding Author:
P. FERNÁNDEZ
pilar.fernandez@ciemat.es
Avda. COMPLUTENSE, nº 22, 28011 MADRID, SPAIN
364
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P1C-I-150
REACTION OF TITANIUM BERYLLIDE
MUNAKATA KENZO, KAWAMURA HIROSHI (1) UCHIDA MUNENORI (2)
(1) Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193, Japan (2) NGK INSULATORS, LTD.
Maegata, Handa, Aichi 475-0825, Japan
Beryllium is considered as one of the candidate materials of the neutron multiplier in the tritium-breeding
blanket. In the blanket of D-T fusion reactors, beryllium would be placed in the high neutron flux and high
temperature environment. Thus, there are some problems related to the application of beryllium as the neutron
multiplier, which are the compatibility with structural materials, the tritium inventory and the reactivity of
beryllium with water vapor and oxygen in the LOCA accident. Titanium beryllides such as Be12Ti are known
to have advantages over beryllium from the perspectives of higher melting point, lower chemical reactivity,
lower swelling and so forth. Thus, these materials are thought to become promising alternatives of beryllium.
Few experimental data related to the subjects described above are, however, available for these materials.
Therefore more experimental studies need to be done to evaluate their performance as the neutron multiplier.
With respect to the reaction with water vapor, beryllium is known to be highly reactive at high temperatures
and under high vapor pressures of water, which is one of the major drawbacks of beryllium. Therefore, in this
work, the authors investigated the reaction of titanium beryllides with water vapor. In the experiments, the
sample disks of a titanium beryllide were placed in a reactor made of quartz. An argon gas with 10,000 ppm of
water vapor was introduced to the reactor, and the temperature of the reactor was raised up to 1000 C. The
concentrations of water vapor and hydrogen in the outlet stream of the reactor were traced with a massspectrometer. The sample was exposed the wet argon gas at 1000 C for about 20 h. However, the chaotic
breakaway reaction, which is known to take place in the case of beryllium, was not observed. The analysis of
the result reveals that the amount of water, which reacts with Be12Ti, is smaller in comparison with beryllium.
After the experiment, the surface of Be12Ti was investigated by means of digital microscope, scanning
electron microscope, x-ray diffraction analysis and so froth. The results of the analysis indicate that the surface
of Be12Ti was oxidized and the surface roughness increased, but the oxidized layer was not thick. It was also
suggested that the major oxide on the surface is BeO. In the presentation, more detailed results of analysis on
the surface will be presented.
Corresponding Author:
MUNAKATA KENZO
kenzo@nucl.kyushu-u.ac.jp
Kyushu University, Interdisciplinary Graduate School of Engineering Science, Kasuga 816-8580, Japan
365
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P1C-I-158
INTEGRAL BENCHMARK EXPERIMENTS ON VANADIUM SPHERES
WITH A CENTRAL 14-MEV NEUTRON SOURCE AND INSIDE A
SPHERICAL CRITICAL ASSEMBLY
MARKOVSKIJ DMITRY, A.I. BLOKHIN (2) D.YU. CHUVILIN(1) A.V. LIVKE (3)
(1) RRC "Kurchatov Institute", 123182 Moscow, Russia (2) SSC "Institute of Physics and Power Engineering", 249020
Obninsk, Russia (3) RFNC "All-Russia Scientific Research Institute of Experimental Physics" (VNIIEF), 607190 Sarov,
Russia
Vanadium is considered as a main component of low-activation structural materials of potential fusion reactors
and other nuclear-power-generating installations of advanced nuclear energy. The given work performed in
frames of ISTC Project #910.B completes the complex of experimental and calculational researches started by
the previous ISTC Project #910-98 and directed to new evaluation of the vanadium nuclear data and their
validation in benchmark-experiments. The new evaluations of V-50 and V51 isotopic data done in IPPE at the
previous stage of the project were now supplemented with the vanadium-element evaluation. The benchmarkexperiments were continued with different neutron sources (14 MeV-neutron generators and fission critical
assembly) and vanadium spheres of two outer diameters: 24 and 34 cm. The largest vanadium sample was
modernized to enable performing in the RRC “KI” the measurements inside the sphere of neutron reaction
rates (threshold detectors and fission foils) and gamma-dose rates (thermoluminiscent detectors) with a 14MeV neutron source. The obtained C/E values for the reaction rates with the use of IPPE data fit the 10%
corridor of experimental uncertainty. For the data of EFF and, especially, of JFF there is some underestimation
by calculation of scattered neutrons, progressing on sphere radius up to ~20 %. Within the limits of
experimental uncertainty of ±20 %, the calculations with all the used data showed consistence to the
experiment. Vanadium nuclear constants at moderate neutron energies were tested in the experiments on
spherical critical assembly with vanadium sphere in its center, performed in VNIIEF (installation FKBN-2M).
The negative reactivity effect of the 24-cm vanadium sphere calculated with the IPPE vanadium nuclear data
agrees with the experiment within the limits of total uncertainty ~10 %, while the calculations with the other
data show underestimation, especially with the JFF data (~20 %). The activation rates at the center of
vanadium sphere with the IPPE data agree with the experiment better than total uncertainty of 10%, except for
63Cu(n,ã)64Cu and 27Al(n,á)24Na activation rates, overestimated by the calculation with the IPPE transport
data and the IRDF-90 dosimetry file by 20%.
Corresponding Author:
MARKOVSKIJ DMITRY
markd@nfi.kiae.ru
Russian Research Centre "Kurchatov Institute", Kurchatov Sq. 1, 123182 Moscow, Russia
366
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P1C-I-163
PRESENT DEVELOPMENT STATUS OF EUROFER AND ODS FOR
APPLICATION IN BLANKET CONCEPTS
RAINER LINDAU, A.MÖSLANG (1) M.RIETH (1) M.KLIMIANKOU (1) E.M.MORRIS (1) A.ALAMO (2)
F.TAVASSOLI (2) C.CAYRON (3) A.LANCHA (4) P.FERNANDEZ (4) N.BALUC (5) R.SCHÄUBLIN (5)
E.DIEGELE (6) J.RENSMAN (7) B.V.D.SCHAAF (7) E.LUCON (8) W.DIETZ (9)
1)FZK Institute for Materials Karlsruhe, D 2)CEA - Saclay, SRMA/SMPX, F 3)CEA - Grenoble, DRT/DTEN/SMP/LS2M, F
4)CIEMAT, E 5)CRPP – EPFL, CH 6)EFDA/CSU, EFDA CSU,D 7)NRG, MM&I,The NL 8)SCK-CEN, B 9)MECS, D
Selected also for oral presentation
O1A-I-163
Within the European Union, the two major breeder blanket concepts presently being developed are the Helium
Cooled Pebble Bed (HCPB), and the Helium Cooled Lithium Lead (HCLL) blankets. For both concepts,
different conceptual designs are being discussed with temperature windows in the range 250-550 C for
conservative approaches based on reduced activation ferritic-martensitic (RAFM) steels, and in the range 250650 C for more advanced versions, taking into account Oxide Dispersion Strengthened (ODS) steels. As a
result of a systematic screening of RAFM-steels in Europe, the 9% CrWVTa alloy EUROFER was specified
and industrial batches of 3.5 and 8.0 tons have been produced with a variety of semi-finished product forms. A
large characterisation program is being performed including microstructural, mechanical and corrosion
experiments. Irradiation programs in materials test reactors have been performed between 60 and 450 C (£15
dpa), and in a fast breeder reactor at 330 C up to 30 dpa. 75 dpa data will become available in about two years.
EUROFER is resistant to high temperature ageing, and the existing creep-rupture data (~30,000 h, 450 - 600
C) indicated long-term stability and predictability. Although irradiated EUROFER specimens have been
examined at present only at 10 dpa and below, it can be stated that irradiation induced hardening, ductility
reduction and fracture toughness degradation are highly superior to irradiated conventional ferritic-martensitic
steels like EM10, HT9, MANET and T91. A replacement of presently considered RAFM steels by suitable
ODS alloys a substantial increase of the operating temperature from ~550 C to about 650 C or even more. This
has been shown by long-term creep rupture tests (£10,000 h between 600 and 700 C) on specimens made of
ODS-EUROFER97. A large material characterisation program, including reactor irradiation to 15, 30, and 70
dpa, is ongoing. A breakthrough has been achieved at FZK in overcoming the poor high temperature ductility
and ductile-to-brittle-transition temperature (DBTT) of first generation RAFM-ODS and commercial ferritic
ODS alloys. Selecting a specific production route for the mechanically alloyed EUROFER-0.3wt% Y2O3
which included rolling and appropriate thermal treatments, DBTT could be shifted from +60 C for hipped
ODS-Eurofer of the first generation to values well below 0 C. A reliable joining technique for ODS and
RAFM steels employing diffusion welding was successfully developed.
Corresponding Author:
RAINER LINDAU
rainer.lindau@imf.fzk.de
Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe, Germany
367
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P1C-I-164
MICROSTRUCTURAL INVESTIGATION, USING SMALL ANGLE
NEUTRON SCATTERING, OF NEUTRON IRRADIATED EUROFER 97
STEEL
COPPOLA ROBERTO, R. LINDAU(1) M. MAGNANI(2) R. P. MAY (3) A. MÖSLANG (1) J. W. RENSMAN(4) B.
VAN DER SCHAAF(4) M. VALLI(2)
(1)FZK, IMF, PO Box 3640, D-76021 Karlsruhe. (2)ENEA-“Clementel“, V. Don Fiammelli 2, 40129 Bologna, Italy. (3)ILL,
6 rue Jules Horowitz, 38042 Grenoble – F. (4)NRG, PO Box 25, 1755 ZG Petten, The Netherlands.
The reduced activation Eurofer97 ferritic/martensitic steel (9Cr1W0.2VTa0.1C) is being extensively
characterized as a European reference in view of its possible use as a structural material in ITER test blanket
modules and in future demonstration reactors. Both its metallurgical properties and its performance under
irradiation are studied, with special attention to fracture toughness, DBTT and tensile properties. This
contribution will present the results of small-angle neutron scattering (SANS) measurements carried out on
Eurofer 97 neutron irradiated with 2.7 dpa, at 60 C and at 300 C. For comparison samples of a laboratory heat
of this same steel, irradiated under the same conditions, have also been investigated. The SANS measurements
have been carried out at the High Flux Reactor of the ILL-Grenoble. SANS data analysis, based on the
comparison of nuclear and magnetic scattering, shows that a significant fraction of defects produced under
irradiation is non-magnetic (He-bubbles, microvoids).Their distribution strongly varies with the irradiation
conditions, namely with increasing irradiation temperature. A consistent increase is observed both in the
average size (up to approximately 10 nm) as well as in the volume fraction. These findings will be discussed
with reference to post-irradiation mechanical tests carried out over the same material (1) and to ongoing
transmission electron microscopy observations. Ref.: (1) J. Rensman et al., J. Nucl. Mat. 307-311 (2002) 250
Corresponding Author:
COPPOLA ROBERTO
coppolar@casaccia.enea.it
ENEA-Casaccia, FIS, CP 2400, 00100 Roma - I
368
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P1C-I-168
EFFECT ON IMPACT TOUGHNESS OF REDUCED OXYGEN
CONTENT IN 316 STEEL POWDER JOINED TO 316 STEEL BY LOW
TEMPERATURE HIP
LIND ANDERS, PEACOCK ALAN
EFDA Close Support Unit, DE-857 48 Garching Germany
During the manufacture of the blanket modules 316L steel powder is simultaneously consolidated and joined
to tubes and blocks of 316 L materials by Hot Isostatic Pressing (HIP). The processing temperature is normally
relatively high and has a detrimental effect on the grain size of the water cooling tubes in the structure and also
on the blocks reducing their strength. This work aims to investigate the effect on the impact toughness and
other mechanical properties of low temperature HIP joints between the consolidated powder and steel. Joining
temperatures of 1020 C and 1060 C were used. It is well known that the surface oxides on the powder particles
will influence the impact toughness in a negative way. At a high HIP temperature the oxides are at least partly
transformed thereby improving the impact toughness. In order to get acceptable mechanical properties of
materials produced at a low HIP temperature the oxygen content on the powder surfaces has to be reduced.
Techniques to reduce the oxygen content of the powder material were therefore applied. The material
properties for low temperature HIP joints of a steel block and consolidated powder with reduced oxygen
content are compared to those for powder with normal oxygen content. The influence on the grain size of the
HIPing temperatures is exhibited.
Corresponding Author:
LIND ANDERS
anders.lind@studsvik.se
Studsvik Nuclear AB, Euratom Association VR, SE 611 82 Nyköping Sweden
369
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P1C-I-178
ENVIRONMENTAL ASSISTED CRACKING OF EUROFER 97 IN
WATER AND PB-LI
VAN DYCK STEVEN, BOSCH RIK-WOUTER
SCK-CEN, Boeretang 200, B2400 Mol, Belgium
Environmental assisted cracking (EAC) of materials is a critical concern for the design of nuclear systems. For
fusion reactor applications, the ferritic-martensitic Eurofer97 steel is considered a prime candidate for building
future power reactors. In the current design concepts, there are two potentially corrosive environments for this
type of materials, namely water at high temperature and liquid lead-lithium eutectic alloys. It is well known
that the occurrence of EAC depends on both the environment and material conditions. For nuclear
applications, both are subject to modifications due to irradiation. In this paper, the results of a research
programme, dealing with stress corrosion cracking (SCC) of Eurofer97 in water are presented. The influence
of water chemistry, temperature and material condition are discussed. The unirradiated Eurofer97 is
susceptible to SCC in oxygenated water with chloride ion contamination, but can be used securely in purified
water (chloride level below 50ppb) at low corrosion potential. Special attention is given to the effect of
irradiation hardening on stress corrosion cracking. Literature data suggest increased susceptibility to SCC in
materials, hardened by low temperature tempering, welding or oxide dispersion strengthening. The stress
corrosion susceptibility of the irradiated material is assessed by slow strain rate tensile testing, before and after
irradiation up to 2.3 dpa. In a second part, the outline for a programme, investigating the interaction between
irradiation hardening and liquid metal embrittlement (LME), is presented. In analogy to SCC, it is expected
that the irradiation hardening will enhance the material's susceptibility to LME. A new test set-up is developed
for SSRT testing of irradiated materials, in contact with liquid Pb-Li eutectic.
Corresponding Author:
VAN DYCK STEVEN
svdyck@sckcen.be
SCK-CEN, Boeretang 200, B2400 Mol, Belgium
370
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P1C-I-179
MEASUREMENT AND ANALYSIS OF RADIOACTIVITY INDUCED IN
YTTRIUM AND LEAD IN FUSION PEAK NEUTRON FIELD
SEIDEL, KLAUS, RANDY EICHIN (1) ROBIN A. FORREST (2) HARTWIG FREIESLEBEN (1) SIEGFRIED
UNHOLZER (1)
(1) TU Dresden, Institut fuer Kern- und Teilchenphysik, D-01062 Dresden, Germany. (2) Euratom/UKAEA Fusion
Association, Culham Science Centre, Abington OX143DB, United Kingdom.
The large fluxes of neutrons produce in the materials of a fusion device during operation radioactivity that is
relevant to operational safety and decommissioning. Radionuclides with a broad range of half-lives have to be
included in the corresponding analyses. The radioactivity with decay times ranging from the order of
magnitude of minutes to weeks is of interest with respect to heat production and shut-down dose rates, whereas
long-term radioactivity determines the waste management. The radioactivity is mainly produced by two
components of the neutron flux spectrum, by thermal neutrons and by the 14-MeV D-T fusion neutrons.
Analyses of the material activation rely on calculations with inventory codes and libraries containing
activation and decay data. To gain trust in the results of such calculations data and codes have to be validated
experimentally. In the present work, the European Activation System (EASY, inventory code FISPACT and
data library EAF) was tested in benchmark experiments on Y and Pb. Y is a constituent of ODS steels to be
used as structural material in first wall and blanket. Pb is an important material in several breeding blanket
concepts. Small pieces of Y and Pb were irradiated in a D-T neutron field. The gamma-radioactivity following
irradiation was measured several times during decay and nuclide activities were derived. For each of the
measured activities the corresponding value was calculated with EASY, and the calculated-to-experimental
value (C/E) was determined. The nuclear reactions producing the activities were analysed too. The C/E
obtained for the individual activities are used for discussing the activation performance and the contact dose
rate of the materials at fusion reactor conditions.
Corresponding Author:
SEIDEL, KLAUS
seidel@physik.phy.tu-dresden.de
Technische Universitaet Dresden, Institut fuer Kern- und Teilchenphysik, D-01062 Dresden, Germany
371
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P1C-I-183
EVALUATION OF NUCLEAR HEATING, TRITIUM BREEDING AND
SHIELDING EFFICIENCY OF THE DEMO HCLL BREEDER BLANKET
JORDANOVA JORDANA, ULRICH FISCHER(2) PAVEL PERESLAVTSEV(2) YVES POITEVIN(3) ANTONELLA LI
PUMA(3) ANTONIO CARDELLA(4) ANTON ADAMSKI(1)
(1)INRNE,1784 Sofia, Bulgaria (2)Forschungszentrum Karlsruhe, Postfach 3640, 76021 Karksruhe, Germany (3)CEA
Saclay, DRN/DMT/SERMA, 91191 Gif-sur-Yvette, France (4)EFDA CSU Garching, Boltzmann str. 2, D 85748 Garching,
Germany
This work summarizes the main results of the neutronics calculations for the modular HCLL (Helium Cooled
Lithium Lead) breeder blanket design of a DEMO-type reactor. Estimates of the tritium production
capabilities, nuclear energy deposition and shielding efficiency to protect the TF coils are presented. Detailed
three-dimensional radiation transport calculations based on the Monte Carlo code MCNP-4C have been
performed to predict the neutronics characteristics of the blanket. A three-dimensional 90 sector model
representing the modular HCLL breeder blanket developed by FZK has been used in calculations. It was
developed by integrating HCLL breeder blanket modules into neutronic model of a DEMO-type reactor
derived from the PPCS study by CEA and FZK [1]. A proper spatial distribution of D-T source neutrons was
assumed in calculations [1]. Coupled neutron/gamma calculations have been carried out using FENDL-2 and
MCPLIB2 cross-section data sets to obtain the spatial dependence of neutron and gamma heating rates. The
nuclear power generation and its spatial distribution, both radial and poloidal, and the nuclear heat production
of the HCLL modules have been estimated. Tritium production rate has been assessed using FENDL-2 crosssections data to calculate the spatial/energy distribution of neutron flux in breeder zone and the response. The
estimated TBR for the reference case (90% tritium enrichment, breeder zone radial thickness of 75 cm, 2 mm
thick W armor of the first wall) amounts to 1.22. The excess of unity achieved ensures tritium selfsufficiency
if the tritium losses, ports and penetrations are taken into account. Several shielding configurations for the
torus mid-plane have been investigated and the resulting radiation load to the super-conducting magnets has
been assessed. It was found that shielding of super-conducting TF coils can be provided by about 100 cm thick
blanket/shielding system, i.e. the integral Epoxy radiation dose is below the design limit of 1.0x107 Gy
(40FPY). [1] Y. Chen, U.Fischer, P.Pereslavtsev, F. Wasastjerna, The EU Power Plant Conceptual Study –
Neutronic Design Analyses for Near Term and Advanced Reactor Models, FZKA 6763, (April 2003)
Corresponding Author:
JORDANOVA JORDANA
jordanaj@inrne.bas.bg
Institute for Nuclear Research and Nuclear Energy, Tzarigradsko chaussee 72, 1784 Sofia, Bulgaria
372
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P1C-I-189
HYDROGEN EFFECTS ON THE TENSILE AND FATIGUE
PROPERTIES OF EUROFER 97
MADAY MARIE-FRANÇOISE,
In the recent European blanket option which considers the use of reduced activation ferritic/martensitic steels
(RAF/MS) partially coated with diffusion barriers and cooled with helium gas, the main source of hydrogen
for structural materials is predicted to consist of in-bulk (n,p) nuclear transmutations. Theoretical evaluations
made on various candidate alloys have shown that depending on the specific blanket design and operating
parameters, internal hydrogen could locally attain steady-state concentrations potentially conducive to
embrittlement phenomena. Critical conditions for hydrogen-induced damage in a number of 9%Cr steels have
been mainly inferred from the results of tensile tests performed by the constant extension rate technique
(CERT). This procedure, while very suitable for the end use of sorting materials in terms of their relative
embrittlement susceptibility, may nevertheless generate either under or over estimates, especially for practical
applications where complex time-dependent stresses are expected to occur. For instance, cyclic loading has
often been found to cause material failure at hydrogen contents which were retained innocuous from CERT
experiments, or to favor specific cracking mechanisms, which did not dominate fracture in tensile tests. Since
safe predictions about hydrogen-material interaction for a reliable design require extensive data-base
development, complementary inputs from independent test techniques reflecting specific ranges of possible
service experience, have their respective relevance for improving the understanding about this issue. This
paper presents the results of an explorative work performed within the frame of the European Fusion Materials
Programme, on Eurofer'97, which actually is the reference RAF/MS of the European strategy. In this study,
CERT and load-controlled fatigue tests have been run on specimens electrochemically pre-saturated with
various amounts of hydrogen, and held under dynamic charging to maintain saturation during testing. Ductility
loss and fatigue lifetime reductions compared to the uncharged specimen conditions were recorded as a
function of hydrogen contents, which were determined immediately after specimen rupture by thermal
desorption in a LECO analyser. The results obtained by the different methods were compared and discussed
with the support of microstructural and fractographic assessments.
Corresponding Author:
MADAY MARIE-FRANÇOISE
francoise.maday@casaccia.enea.it
ENEA-CR Casaccia, Via Anguillarese 301, 00060 S.Maria di Galeria, Rome, Italy
373
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P1C-I-191
THE HELIUM COOLED LITHIUM LEAD BLANKET TEST PROPOSAL
IN ITER AND REQUIREMENTS ON TEST BLANKET MODULES
INSTRUMENTATION
LI PUMA ANTONELLA, POITEVIN YVES(1) GIANCARLI LUCIANO(2) RAMPAL GILLES(1) FARABOLINI
WILFRID(1)
(1) CEA/Saclay, DEN/DM2S F-91191 Gif-sur-Yvette, France (2) CEA/Saclay, DEN/CPT F-91191 Gif-sur-Yvette, France
One of the missions of ITER is "to test tritium breeding module concepts that would lead in a future reactor to
tritium self-sufficiency, the extraction of high grade heat and electricity production". This requires the
development of test modules based on corresponding DEMO blanket design but using appropriate engineering
scaling, to allow the achievement of the different test objectives in spite of the differences in operating
conditions between DEMO and ITER. The Helium Cooled Lithium Lead (HCLL) is one of the two European
breeding blanket lines chosen to be further developed with the aim of manufacturing a Test Blanket Module
(TBM) for ITER. In this paper, a detailed TBMs testing sequence is proposed. The proposal envisages
different test mock-ups or modules, which are optimized for single or combined effects and whose design
makes large use of engineering scaling for compensating the significant difference between the testing
conditions and those expected in DEMO (e.g., temperature gradients, pulsed T-production, coolant flow rate).
The objective is to demonstrate the validity of the different components and the feasibility of measure
instruments and results interpretation. In order to take the maximum benefit from all the ITER phases and
related operating conditions, including the initial H-H phase, it is foreseen to test 4 types of TBM, for which
the design rational and objectives of tests are presented and discussed. The choice of each type of module is
defined in a way that it will provide a progressive qualification of the HCLL blanket line up to an integral
demonstration in the final TBM (TBM-Integral). Preliminary requirements on necessary instrumentation are
identified for each envisaged TBM. The EU proposal will have to be discussed with the other ITER partners
interested in the development of blanket lines using eutectic Pb-15.7Li as breeder material in order to launch
international collaborative actions.
Corresponding Author:
LI PUMA ANTONELLA
alipuma@cea.fr
CEA/Saclay DEN/DM2S/SERMA/LCA, F-91191 Gif sur Yvette CEDEX France
374
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P1C-I-196
NUMERICAL AND EXPERIMENTAL STUDY ON TIME-DEPENDENT
THERMOMCHANIC DEFORMATION OF CERAMIC BREEDER
PEBBLE BEDS
YING, ALICE, AN, JOHN ABDOU, MOHAMED
Significant effort is underway to understand and quantify time-dependent thermomechanic behavior of
heterogeneous systems such as ceramic breeder pebble beds. Unlike solid material, stress distribution in such a
system is highly concentrated in a time-dependent local zone of contact area. Initially, deformation can be
locally intense due to stress concentration, but it can become relaxed as the contact area grows and stress
magnitude is reduced. Thus, the deformation that is exhibited in a pebble bed system as time evolves can
derive from different mechanisms, including power law and diffusion flow. It is critical that a predictive tool is
developed in order to ensure that blanket system performance is not endangered or limited. This paper presents
the progress on the development of this predictive capability based on a discrete numerical scheme as well as a
finite element approach. Calculations have shown that the stress magnitude at the contact can be as high as
several hundred MPa even though its measurable global value appears at only a few MPa. Without taking into
account this insight, previous analysis based on a Coble creep mechanism fails to predict the initial creep rate
and shows a significantly lower value as compared to experimental results. This discrepancy has been
minimized when a power law like creep mechanism with a stress exponent factor of 1.86 is considered.
However, in view of the interwoven complexity embedded in the phenomenon, a parametric study is
conducted to identify the most important dominating physical parameters while their impacts are quantified.
The other important characteristics for time-dependent deformation consideration are related to stress
relaxation time. Stress build-up in the system, via either differential thermal expansion or irradiation swelling,
must decay fast enough to prevent adverse operating conditions such as an increase in thermal resistance as the
result of formation of a gap at the interface. Analysis based on a diffusional relaxation mechanism has shown a
stress reduction by one order of magnitude in about 20 hours. Further comprehension of the implications of the
time magnitude of stress evolution as well as interaction between different stress regimes with respect to
deformation propagation is needed. Experiments are being carried out in order to provide data that can validate
the model developed and to augment our knowledge on this issue.
Corresponding Author:
YING, ALICE
ying@fusion.ucla.edu
Mechanical and Aerospace Engineering Dept., UCLA, Los Angeles, CA 90095-1597, USA
375
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P1C-I-197
HYDROGEN ISOTOPE DISTRIBUTIONS AND RETENTION IN THE
INNER DIVERTOR TILE OF JT-60U
YASUHISA OYA, Y.HIROHATA(1) K.SHIBAHARA(2) M.OYAIDZU(3) T.ARAI(4) K.MASAKI(4) Y.GOTOH(4)
K.OKUNO(3) N.MIYA(4) T.TANABE(2) T.HINO(1) S.TANAKA(5)
(1)Hokkaido University, Sapporo, Japan (2)Nagoya University, Nagoya, Japan (3)Shizuoka University, Shizuoka, Japan
(4)Japan Atomic Energy Research Institute, Ibaraki, Japan (5)The University of Tokyo, Tokyo, Japan
Tritium retention in carbon materials is one of the most critical issues. According to recent retention studies of
hydrogen isotopes (H, D, T) in plasma facing carbon tiles used in JT-60U, most of tritium was not codeposited
with carbon but implanted into the depth of micrometer range, while H and D were retained very near the
surface. However, H and D distributions were quite different with each other showing much shallower depth
profiles for H. Because most of the deuterium once retained in the near surface region was replaced by
hydrogen during the final HH discharge and/or exposure to the atmosphere. This indicates that, in order to
estimate tritium retention and/or understand tritium behavior in a fusion reactor, one should study behavior of
all hydrogen isotopes in tokamaks in terms of deposition, temperatures of tiles and discharge history. In this
study, the profiles of hydrogen and deuterium retained in the graphite tiles placed in the divertor region of JT60U were analyzed by secondary ion mass spectroscopy and thermal desorption spectroscopy. It was found
that the hydrogen (H+D) retention in shallower region of the outer divertor was higher than that of the inner
divertor, which is opposite to the results in JET and other tokamaks that the most of hydrogen was retained in
the deposition layers and hence the hydrogen retention in the inner divertor (where is deposition dominated)
was higher. However, the high hydrogen (H+D) retention in the outer divertor was limited only very near
surface region and decreased quickly toward the deeper region. In contrast, hydrogen retention in the inner
divertor was rather small at the very near surface because of poor thermal contact of the re-deposition layer to
the substrate resulting in high surface temperature but retained uniformly within the depth of more than 1.7
micron. This difference well corresponds to the observation that the outer divertor tiles were mostly eroded
and the surface temperature of the tile during the discharge was high (over 400 C), while the inner divertor was
covered by the thick re-deposition layer. Nevertheless, hydrogen (H+D) retention in the deposited layers was
much less compared with that in JET. This indicates that the tritium retention in deposited layers at the higher
temperature region would not have large contribution. In other words, hydrogen isotope retention could be
reduced by controlling the temperature of the re-deposition area.
Corresponding Author:
YASUHISA OYA
yoya@ric.u-tokyo.ac.jp
Radioisotope Center, The University of Tokyo, 2-11-16, Yayoi, Bunkyo-ku, Tokyo, 113-0032 Japan
376
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P1C-I-200
CRYSTAL STRUCTURE OF LI2TIO3 WITH SOME DIFFERENT OXIDE
ADDITIVES
TSUYOSHI HOSHINO, KUNIHIKO TSUCHIYA(1), KIMIO HAYASHI(1), TAKAYUKI TERAI(2), SATORU
TANAKA(3), YOICHI TAKAHASHI(4)
(1)Oarai Research Establishment, Japan Atomic Energy Research Institute, Japan (2)Nuclear Eng. Research Laboratory,
The Univ. of Tokyo, Japan (3)Graduate School of Eng., The Univ. of Tokyo, Japan (4)Department of Applied Chem., Chuo
Univ., Japan
Li2TiO3 is one of the most promising candidates for solid breeder materials. In order to control the grain
growth at the time of high temperature use, the development of Li2TiO3 which has oxide additive is needed.
Furthermore, since Li2TiO3 is reduced in hydrogen atmosphere, it is important to investigate the reduction
characteristic with oxide addition. In the present paper, the structure and the non-stoichiometry of Li2TiO3
added with some different oxides have been extensively investigated by means of thermogravimetry, X-ray
diffraction (XRD) and so on. In the case of the Li2TiO3 samples used in the present study, Li2TiO3 powder
and oxide powders (CaO, ZrO2, Sc2O3) of small quantity powders were mixed in the proportions
corresponding to the molecular ratio of oxide/Li2TiO3; CaO/Li2TiO3 = 0.20%, ZrO2/Li2TiO3 = 0.44% or
Sc2O3/Li2TiO3 = 0.40%. These samples are designated as Ca-Li2TiO3, Zr-Li2TiO3, and Sc-Li2TiO3,
respectively. XRD measurement showed that the structure of Li2TiO3 which has the oxide additives changed
as follows. Ca-Li2TiO3 : Li2TiO3 + xCaO -> Li2Ti1-xO3 + xCaTiO3 Zr-Li2TiO3 : Li2TiO3 + xZrO2 ->
Li2TiO3 + xZrO2 Sc-Li2TiO3 : 1-xLi2TiO3 + xSc2O3 -> Li2-2xTi1-xScxO3-3/2x Ca-Li2TiO3 and ZrLi2TiO3 exist as double phases, and Sc-Li2TiO3 exists as a single phase and then this reaction product has
influence on the grain growth. Especially, the chemical formula of Li2TiO3 (Li2O/TiO2 = 1) in Ca-Li2TiO3
changes to non-stoichiometric composition, Li2Ti1-xO3 (Li2O/TiO2 > 1), with oxide addition. In
thermogravimetry, the mass of Li2TiO3 was found to decrease with time in the hydrogen atmosphere, then to
increase after the change of the atmosphere from hydrogen to oxygen. The color was observed to change from
white to thin gray or a light-brown under the hydrogen atmosphere. This color-change indicates that the
oxygen content of the sample decreased, suggesting the change from Ti4+ to Ti3+. Further, Ca-Li2TiO3 has
fewer oxygen defects than the other kinds of Li2TiO3. Ca-Li2TiO3 has the smallest mass of TiO2 in Li2TiO3,
so that the order of oxygen defects was as follows, Ca-Li2TiO3 < Li2TiO3 = Zr-Li2TiO3 < Sc-Li2TiO3
according to the order of molar ratio Li2O/TiO2 of the samples. The overall results suggest that the oxide
additives are able to control not only the growth of the grain size but also the amount of oxygen defects. Thus,
the validity of oxide addition to Li2TiO3 has been confirmed.
Corresponding Author:
TSUYOSHI HOSHINO
hoshino@sky.biglobe.ne.jp
Blanket Irradiation and Analysis Laboratory, Department of JMTR Project, Oarai Research Establishment,
Japan Atomic Energy Research Institute, 3607, Narita-cho, Oarai-machi, Higashi Ibaraki-gun, Ibaraki, 3111394, Japan
377
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P1C-I-205
EVALUATION OF INSULATING PROPERTY OF CERAMIC
MATERIALS FOR V/LI BLANKET SYSTEM UNDER FISSION
REACTOR IRRADIATION
TERUYA TANAKA, AKIHIRO SUZUKI(1) TAKEO MUROGA (2) TATSUO SHIKAMA (3) MINORU NARUI (3) BUN
TSUCHIYA (3)
(1)University of Tokyo,2-22 Shirakata-Shirane, Tokai-mura, Naka-gun, Ibaraki 319-1188, JAPAN (2)National Institute for
Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, JAPAN (3)Tohoku University, 2145-2 Narita-cho, Oarai, Ibaraki 3111313, Japan
The development of electrical insulating coating has been conducted for reduction of MHD effects in Li/V
blanket system. According to previous studies, nonconventional ceramic materials such as Er2O3, Y2O3,
CaZrO3 and others have been selected as candidates mainly because of their viability in highly corrosive
liquid Li. For the purpose of examining the insulating properties under radiation environment, low flux DT
neutron irradiation of up to ~0.1 Gy/s was performed previously on bulk and coating specimens of the
candidate materials. Extrapolation of flux dependence of the radiation induced conductivities indicated that the
materials are expected to satisfy required insulating performance under radiations at fusion blanket. It was
necessary to evaluate the conductivities at higher dose rate for verification of the extrapolation. In the present
study, insulating properties of some candidates were examined at higher dose rate of up to several tens Gy/s
using a fission reactor. Irradiation experiment on bulk specimens of Er2O3, Y2O3 and CaZrO3 was performed
at JMTR (Japan Material Testing Reactor) of JAERI. The dimension of the disc specimens made by sintering
method was 10 mm in diameter and 1 mm in thickness. Pt electrodes were made by sputter deposition on both
sides of the specimens. The electrical conductivities before irradiation were lower than 10-12 S/m. The
specimens installed in an irradiation capsule were connected to a voltage supply and an electrometer via
mineral-insulated cables. The insulating properties were examined by in-situ measurement of currents flowing
through the specimens. Dose rates and temperatures were 8 - 50 Gy/S and 303 - 343 K, respectively. While
radiation induced currents were changed coincidently with bias voltages, non-ohmic characteristics was
observed in the I-V (current-voltage) curves. It might be due to ionization of helium gas for temperature
control in irradiation capsule. Degradation of insulating properties was evaluated as radiation induced
conductivities (RIC) from the maximum currents for bias voltage of +100 V. The values of RIC were 5.2 x 109 S/m for ~8 G/s (Y2O3), 1.8 x 10-8 S/m for ~40 Gy/s (CaZrO3) and 2.1 x 10-8 S/m for ~50 Gy/s (Er2O3).
The data were shown to be close to values predicted by the extrapolation of the low flux DT neutron
irradiation data.
Corresponding Author:
TERUYA TANAKA
teru@nifs.ac.jp
Fusion Engineering Research Center, National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292,
JAPAN
378
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P1C-I-214
EVALUATION OF HYDROGEN ISOTOPE RETENTION IN BE12TI AS
NEUTRON MULTIPLIER OF FUSION REACTOR
NAOAKI YOSHIDA(1), HIROTOMO IWAKIRI(1) KAZUFUMI YASUNAGA(1) MUNENORI UCHIDA(2) HIROSHI
KAWAMURA(3)
(1)Kyushu University, 6-1 Kasugakoen, Kasuga, Fukuoka 816-8580, Japan (2)NGK Insulators. LTD. Maegata, Handa, Aichi,
475-0825, Japan (3)Japan Atomic Energy Research Institute, Naka, Ibaraki 311-0194, Japan
In the design of a fusion reactor, Be is a reference material for the neutron multiplier of the blanket. However,
it is pointed out that high reactivity and large swelling under high dose neutron irradiation will be serious
disadvantages of this material. Recently, beryllides such as Be12Ti attract a great deal attention as potential
candidates for advanced neutron multipliers from the viewpoints of high melting point, high Be content, fast
decay of gamma dose rate and good chemical stability. Low retention of tritium is also necessary caracteristics
as a neutron multiplier. In the present study, therefore, behavior of implanted deuterium was evaluated for
Be12Ti. Formation of radiation damage, which essentially controls the behavior of hydrogen isotopes, was
also examined to understand the atomistic mechanism of hydrogen isotopes retention under neutron irradiation
condition. Damage evolution by low energy deuterium ion irradiation and the thermal desorption of the
deuterium were examined by using TEM and TDS, respectively. High energy heavy ion irradiations were also
carried out to reveal the fundamental aspect of radiation damage of the material. As an advantage feature,
formation of radiation induced defects such as dislocation loops and voids is very inactive in this material, but
as a very special case deuterium bubbles are formed in wide temperature range by low energy deuterium ion
irradiation, where the apa rate of deuterium is comparable with dpa rate. TDS measurements for the deuterium
ion irradiated specimens showed that total retention of deuterium in Be12Ti is much lower than that of Be at
the wide temperature range and most of them are released up to 500K due to rather weak trapping. Once the
bubbles are formed they also act as trapping sites for deuterium. Annealing experiments under TEM
observation indicate that the bubbles can thermally migrate above 573K and diminish around 1273K by
releasing deuterium gas retained inside. It is worthy of notice that the amount of deuterium retained in the
bubbles in Be12Ti is more than two orders of magnitude lower than that in bubbles of Be. The present results
of low damage accumulation and low efficiency of deuterium retention in bubbles indicate that Be12Ti must
be superior as neutron multiplier to Be in respect to tritium retention.
Corresponding Author:
NAOAKI YOSHIDA(1)
yoshida@riam.kyushu-u.ac.jp
6-1 Kasugakoen, Kasuga, Fukuoka 816-8580, Japan
379
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P1C-I-215
MECHANICAL PROPERTIES OF WELDAMENT USING IRRADIATED
STAINLESS STEEL FOR BLANKET
YAMADA HIROKAZU, HIROSHI KAWAMURA(1) WATARU KOHNO(2) YASUO MORISHIMA(3)
1: JAERI, Naka-machi, Naka-gun, lbaraki-ken, 311-0193, Japan 2: Toshiba Corporation, Tsurumi-ku Yokohama-shi
Kanagawa-ken, 235-8523 Japan 3: Nippon Nuclear Fuel Development Co., Ltd., Oarai-machi Higashiibaraki-gun Ibarakiken, 311-1313 Japan
At the installation of a new blanket of ITER, a cooling water pipe from backside of blanket is jointed to an
irradiated cooling water pipe by a laser welding method. On the welding of irradiated material, joining
technology with irradiated structural materials has been developed in the way of tungsten inert gas (TIG)
welding or laser welding by recent studies. In the ITER, the load by cooling water flow vibration, force by
plasma disruption and deadweight of blanket act to the cooling water pipe from backside of blanket. Therefore,
the mechanical properties of welding joint by the laser welding should be evaluated for installation of a new
blanket. In this study, the bending properties of welding joint of irradiated material and un-irradiated material
(irradiated/un-irradiated joints) were investigated using SS316LN-IG, which is the candidate material for the
cooling pipe of ITER. Additionally, the bending properties of welding joints of irradiated material and
irradiated material (irradiated/irradiated joints), un-irradiated material and un-irradiated material (unirradiated/un-irradiated joints) were clarified for the effect of material combination. The irradiated material for
joints was obtained by irradiation in the Japan Materials Testing Reactor (JMTR), and the irradiation damage
was about 0.3 dpa. The helium contents of irradiated material was about 3 appm. The results of this study
showed that the bending position of joints using un-irradiated material was un-irradiated part and that the
bending position of irradiated/irradiated joints was fusion area or HAZ (heat affected zone). Although the
bending position of joints was different bor the combination pattern between irradiated and un-irradiated
materials, the bending strength of joint was almost same. Additionally, it is confirmed that bending strength
did not depend on the combination pattern between the irradiated and un-irradiated materials, nor on the
relationship between the heat input direction and the bending load direction. At the conference, bending
properties and the effect of hardening by irradiation will be shown.
Corresponding Author:
YAMADA HIROKAZU
yamada@oarai.jaeri.go.jp
3607 Narita-cho Oarai-machi Higashiibaraki-gun Ibaraki-ken 311-1394
380
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P1C-I-224
AB-INITIO VALUES OF THE HE SIEVERT´S CONSTANT IN LIQUID LI
L. A. SEDANO, A. HASSANEIN (+) J. SANZ (++)
(+) ANL, 9700 South Class Av., Argonne, IL (USA), hassanein@anl.gov (++) ETSII-UNED, c/ Juan del Rosal, 12, 28040
Madrid (E) , jsanz@ind.uned.es
Even though it is commonly accepted that (He), as noble gases, is inert and thus insoluble in LMB,
experimental evidences exist on the fact that (He) solubility Henry’s constant: KH is not an immeasurably
small quantity. Such very low KH imposes severe restrictions to classical methods for the measurement of the
(He) solution magnitudes in LMB finally driving to a deep lack of fundamental data on (He) vs. LMB
interaction. As a consequence, the role of (He) in fusion LM systems is commonly disregarded and (He)cavitation neglected as a key issue for the design of fusion systems. Available constitutional data (structure’s
factors) for monotectic Li has been surveyed in order estimate (He) partial molar heat of solution from Li
radial distribution functions and inter-atomic potentials through Neff-Macquire´s statistical-mechanics models.
Embedding Atom Methods (EAM) are developed to have additional predictions of the partial molar heat of
solution (-Hs) by direct simulation of metal cohesion, He-metal and He-He interaction. He versus Li
embedding potentials are obtained by fitting EAM to available quantum-mechanics results. NVT-canonical
simulations, N = 1000 Li atoms, #V = 10, #T=6 are carried out in a 3D periodical cell. From an initial Li
crystalline configuration the transition to a liquid Li state having the static liquid structure’s factors reported in
literature is simulated ab-initio for computation times large enough so that reliable pressures of the systems
could be evaluated and the fluctuations in boundary pressures minimized. Once Li liquid structure obtained, 10
(He) atoms are added, one by one, to the Li system. Parallel lines for each case, with slopes clearly
independent of number of (He) atoms in the system, are obtained for energy vs. pressure and volume vs.
pressures at given temperature. Average differences between two adjacent parallels at zero pressure, once
kinetic energy of the system discounted, represents the energy gained by an (He) atom when added to the Li
system, i.e.: - Hs. For temperatures between 600 and 900 ºC, the values for the Henry’s constant of (He) in
liquid lithium Li obtained from ab-initio computations range from 8.10-14 to 10-13 at.fr. Pa-1.
Corresponding Author:
L. A. SEDANO
luis.sedano@ciemat.es
Assoc. EURATOM-CIEMAT for Fusion, Bd. 43 P0.04 Avda. Complutense 22, E-28040 Madrid
381
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P1C-I-237
OUT-OF-PILE TRITIUM RELEASE PROPERTY CORRELATIONS FOR
LI-DEPLETED LI2TIO3 AND LI4TI5O12 CERAMICS. EFFECTS OF
REDUCTION-ANNEALING TREATMENTS
CASADIO SERGIO (1),, J.D LULEWICZ(2), J.G. VAN DER LAAN (3), N ROUX (2), A.J. MAGIELSEN (3), M.P.
STIJKEL (3) AND C. ALVANI (1)
1-ENEA-Casaccia, V. Anguillarese, 301, 00060 S.M. di Galeria, Rome, Italy. 2-CEA-Saclay, .F-91191 Gif-sur-Yvette,
CEDEX Saclay, France 3-NRG-Petten, P.O. Box 25, 1755 ZG Petten, The Netherlands
Out-of-pile measurements of tritium release rate from shortly irradiated lithium-titanate ceramics purged by
He+0.1%H2 gas were performed by Temperature Programmed Desorption (TPD) methods and by
distinguishing HT from HTO molecular species. Typical TPD spectra for pebbles and pellets of near pure
Li2TiO3(monoclinic phase), of pure Li4Ti5O12 (spinel phase) and of Li-depleted metatitanate, in which the
spinel phase was found to accompany the m-Li2TiO3. The reduction of the Li/Ti atom ratio to values less than
the stoichiometric one (2.00 in this kind of breeder material) could either be designed in the fabrication route,
or to occur because of lithium evaporation at high temperature and, of course, because the Li-6 transmutation
under irradiation in operating tritium breeding conditions. Although the almost complex patterns of the
resulting TPD spectra, no significant difference was found in the behavior of the two phases when HTO-TPD
signal was concerning. That is due to the water adsorbed on both the "as received" specimens. Nevertheless a
difference was detected for the less intense signal (HT-TPD-spectra)due to the release rate of HT or T2
elemental (reduced) tritium species . The spinel phase specimen showed the maximum release rate (HT-TPDpeak) at a temperature higher than that of the HTO-TPD-peak (while this last was very close to that of the Limeta-titanate). The spectra elaboration was performed by assuming thermally activated Arrhenius type firstorder desorption kinetics hold for all the distinguishable desorption sites (peaks). On the light of this result a
reviewing of previously published data provides a reasonable explanation about the nature tritium trapping
sites observed to arise on reduction-annealed HCPB reference Li-titanate pebbles. They should be due to the
presence of the spinel phase (in its reduced form) at the grain boundaries. Presenting author: Sergio Casadio
ENEA Casaccia, Via Anguillarese, 301, 00060 Rome (Italy) sergio.casadio@casaccia.enea.it Fax:
+39.06.3048.3327 Tel.: +39.06.3048.3246
Corresponding Author:
CASADIO SERGIO (1),
sergio.casadio@casaccia.enea.it
ENEA, CR Casaccia, via Anguillarese, 301, 00060 S.M. di Galeria, Rome, Italy
382
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P1T-I-238
MAGNETOHYDRODYNAMIC PRESSURE-DRIVEN FLOWS IN THE
HCLL BLANKET
BÜHLER LEO,
In currently investigated Helium-Cooled Lithium Lead (HCLL) blankets helium is used for entire heat
removal. For that reason the velocities of the liquid metal can be kept as small as possible to avoid high MHD
pressure drop. However, the velocities should be high enough to ensure sufficient convective transport of
dissolved species like tritium form the blanket to the purification systems. The present work focuses on
pressure drop and flow pattern in typical elements of HCLL breeder units. The magnetohydrodynamic flow is
analysed by a combined asymptotic-numeric approach, valid especially for slow flows in very strong magnetic
fields. Solutions are obtained by using boundary fitted coordinates that allow an efficient description of almost
any arbitrary geometry. Since in the current design the feeding pipes are relatively small compared with the
dimensions of the breeding boxes the velocities there may reach such values that the major fraction of the total
pressure drop arises rather in these ducts than in the breeder boxes. Three-dimensional reorganisation of the
flow at expansions and contractions causes additional pressure drop. The total pressure drop of the MHD flow
through an entire breeder unit is estimated for different flow rates.
Corresponding Author:
BÜHLER LEO
leo.buehler@iket.fzk.de
Forschungszentrum Karlsruhe, Postfach 3640, 76021 Karlsruhe, Germany
383
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P1T-I-242
LIQUID LITHIUM AS THE COOLANT OF THE IFMIF LOOP
LOGINOV NIKOLAY, M.N.ARNOLDOV, H.NAKAMURA, A.S.MIKHEYEV, V.A.MOROZOV, V.M.CHERNOV
IPPE, Bondarenko sq. 1, Obninsk, Kaluga Region, 249033, Russia
The feature of lithium technology in IFMIF loop is considered in the report. Lithium impurities are
enumerating and possible demands to their contents are discussing. The features of lithium purification and
impurity contents measurement methods are discussed. Some information about lithium test facility being
under construction in IPPE is giving. On this facility hydrodynamic and technological problems of IFMIF
target will be studied. Some common aspects exist between lithium IFMIF loop technology and the sodium
cooled fast reactors’ loops. Physical and chemical properties of sodium and lithium are different, nevertheless
several procedures of sodium engineering could be applicable for IFMIF lithium technology. But, essential
differences both thermo-hydraulic parameters and design of a specific facility influence on the requirements to
coolant quality and technology of its handling. The differences are related mainly with impurities composition
and their contribution into the coolant behavior. Liquid-metal coolants impurity monitoring reasons are as
follows: 1. The impurities influence on structure material corrosion. 2. Partial or full blockage of the pipings’
and reactor core’s flow cross-sections by sediments. 3. Blocking of heat exchanger surfaces and heat exchange
process derating due to the impurities build-up. 4. Radiation environment degradation. Lithium purification
could be done by technology similar to sodium and sodium-potassium coolants technology well-proved both
of commercial nuclear power plants and for R&D lithium coolant facilities (cold and hot (i.e. chemical) traps,
distillation, filtration etc.). Relatively low lithium temperature of the IFMIF loop makes hot trap using doubtful
process. Development of the methods and devices for lithium purity monitoring will be carried out in two
stages. The first stage is reviewing and consideration of methods and devices for monitoring the lithium purity
in existing experimental Lithium Test Facility (LTF) under specific IFMIF loop peculiarities. Second stage
involves methods and devices for full scale IFMIF loop. The main methods and devices for purity monitoring
are: plug indicator, sampler-distiller, sampler for chemical analysis, electric resistance indicator, diffusion cells
for hydrogen, carbon, nitrogen, electrochemical cells for oxygen and sodium, radioactivation procedures,
Equilibrium Standard Specimens.
Corresponding Author:
LOGINOV NIKOLAY
loginov@ippe.obninsk.ru
IPPE, Bondarenko sq. 1, Obninsk 249033, Kaluga region, Russia
384
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HCLL TBM FOR ITER – DESIGN STUDIES
P1T-I-252
RAMPAL GILLES, Y. POITEVIN (1) A. LI-PUMA (1) E. RIGAL (2) J. SZCZEPANSKI (3) CÉCILE BOUDOT (4)
(1) CEA Saclay, DEN/DM2S, F-91191 Gif-sur-Yvette, France (2) CEA Grenoble, DTEN/S3ME, F-38054 Grenoble, France
(3) CONCEPT-21, F-91420 Morangis, France (4) FRAMATOME/ANP, Centre technique, F-71205 Le Creusot, France
Within the scope of blanket module design for DEMO for an Helium-Cooled Lithium-Lead (HCLL) Blanket,
test blanket modules (TBM), derived from the DEMO models, have to be designed and inserted in ITER for
testing. Depending on the testing objectives, several TBM have to be designed and manufactured. This paper
describes the specific design studies performed on the so-called “Integral TBM (I-TBM)” which includes all
main features of the corresponding DEMO blanket module, such as design architectures, EUROFER structural
material, functional parameters (e.g., He pressure of 8 MPa), and all relevant manufacturing process. In
parallel with papers on HCLL blankets test strategy and programs, the aim is to present the analyses and
calculations that have lead to the actual geometry. The HCLL I-TBM overall design is very similar to a typical
DEMO module (“act-alike” mock-up). It consists of a steel box directly cooled by He flowing in internal
channels. Welded into the box is a stiffening grid of radial-toroidal and radial-poloidal plates; each grid plate is
cooled by He flowing in internal channels. The Pb-17Li breeder fills the space inside the box and slowly recirculates throughout the box for allowing external T-extraction. The cooling is radial; all the large He
manifolds are in the rear part of the blanket. The I-TBM has been designed to allow the testing of the concept
functionalities, such as helium cooling, magneto-hydrodynamics, tritium permeation, thermo-mechanical
withstanding. Detailed design has also to identify specific needs and to report information to industrials in
terms of fabrication and assembly, such as cooling plates and walls fabrication, welding techniques and their
implication on the mounting sequence. This collaboration with industry has lead to improve the design
regarding to the manufacturing, and to propose a coherent mounting sequence. The TBM design also includes
the external helium and lithium-lead manifolds, and the module attachment to the ITER frame which has to
withstand disruptions. Several calculations have been performed to optimize functional parameters (thermohydraulic evaluations on the helium cooling system, thermal calculations of module structure) and geometry
parameters (mechanical and thermo-mechanical calculations of module structure under normal and accidental
conditions, mechanical calculations of the attachments). Some of the dimensional optimizations were implied
by the manufacturing requirements.
Corresponding Author:
RAMPAL GILLES
gilles.rampal@cea.fr
CEA Saclay, DEN/DM2S, F-91191 Gif-sur-Yvette, France
385
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P1T-I-254
INTERNATIONAL COMPARISON OF MEASURING TECHNIQUES OF
TRITIUM PRODUCTION FOR FUSION NEUTRONICS EXPERIMENTS
BATISTONI PAOLA, M. ANGELONE(1), P. CARCONI(2), K. OCHIAI(3), M. PILLON(1), I. SCHÄFER(4), K.
SEIDEL(5), Y. VERZILOV(3), G. ZAPPA(2)
(1) Associazione Euratom-ENEA sulla Fusione, 00044 Frascati, Italy (2) ENEA, C.R. Casaccia,Italy (3) JAERI, Naka Fusion
Research Estab., Ibaraki-Ken, Japan (4) VKTA Rossendorf, 01314 Dresden, Germany (5) TU Dresden, 01062 Dresden,
Germany
Tritium self-sufficiency is a key issue in the development of fusion reactor based on DT burning. Two Test
Blanket Modules (TBMs), i.e. the Helium-cooled Pebble Bed (HCPB) and the Helium-cooled Lithium-Lead
(HCLL), are currently developed in Europe to be tested in ITER. Major objectives of the TBM tests in ITER
from the neutronics point of view are to demonstrate the tritium breeding performance of the breeder blanket
concepts and to check and validate the capability of the neutronics codes and data to predict the nuclear
responses in the TBM with sufficiently high accuracy. This will allow the computational tools and data to be
applied with confidence in design calculations for fusion power plants. The assessment of the output of nuclear
tests on TBMs will be done in terms of the comparison (C/E ratios) between results of measured tritium
production ratio (TPR, E) in TBMs and the calculated predictions (C) taking into account the uncertainties
both on measurements and calculations. C/E deviations from unity within the total uncertainties will be
regarded as experimental confirmation of numerical predictions, while larger deviations will lead to the
conclusion of non-reliability of nuclear data or numerical tools. The value of the test output will depend on the
quality of both the experimental and numerical tools used, i.e. on the narrowness of uncertainties. Therefore,
the efficient exploitation of TBMs tests on ITER requires that, the computational tools and nuclear data are
made available and validated by means of integral experiments on TBM mock-ups irradiated with neutrons of
appropriate spectral distribution. For this purpose, a neutronics experiment has been launched in the European
Fusion Technology Program on a mock up of one of the European TBMs (the HCPB) with the objective to
reduce uncertainties in the TPR predictions. The objective of the experiment requires the development of both
accurate experimental techniques for tritium measurements, with as low as possible uncertainties, and of
sensitivity/uncertainty analysis tools to provide uncertainties on calculations due to nuclear data. The present
paper describes results of the international benchmark comparison on tritium measurement techniques, carried
out in collaboration between ENEA, TU Dresden and JAERI, with the purpose to prepare accurate
measurement procedures to be used later in the TBM experiment and to carefully assess the associated
uncertainties.
Corresponding Author:
BATISTONI PAOLA
batistoni@frascati.enea.it
Associazione Euratom-ENEA sulla Fusione , ENEA C.R. Frascati, Via E. Fermi 45, I-00040 FRASCATI, ITALY
386
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P1T-I-257
PEBBLE BED THERMAL-MECHANICAL THEORETICAL MODEL:
APPLICATION AT THE GEOMETRY OF TEST BLANKET MODULE
OF ITER-FEAT NUCLEAR FUSION REACTOR
AQUARO DONATO, NICOLA ZACCARI PHD STUDENT TEL. 050/836689
Via Dio ti salvi, n 1 Dipartimento di ingegneria meccanica, nucleare Università di Pisa
This paper deals with a research activity performed in the framework of ITER – FEAT project. Several
nuclear, thermo-hydraulic and thermo-mechanical problems need a better knowledge in order to
technologically develop these pebble bed blankets. These problems concern mainly the theoretical modelling
of the thermal-mechanical behaviour of the pebble bed. Presently a comprehensive theory of the thermalmechanical behaviour of the pebble bed has not been developed yet. At the University of Pisa University a
research activity is carrying out on the breeding blanket of nuclear fusion reactors. In particular a theoretical
and numerical model to simulate the thermal-mechanical behaviour of the pebble bed is being developed. This
model, based on the analysis of the elastic plastic interactions between the pebbles, tries to combine the
advantages of the discrete and continuous methods without their main drawbacks. The mechanical (stiffness )
and thermal (conductivity) characteristics of the bed is deduced considering it made of regular lattices of
pebbles (simple cubic , cubic centred body and cubic centred face) and applying the combination rules of
series or parallel systems. The combination of these regular lattices depends on a packing factor ( the ratio
between the bed density and the density of the pebble material). Moreover, with this model, it is possible to
simulate the steady state creep under static pressure and high temperature. In order to simulate complex
geometries as the module of the breeding blanket of the nuclear fusion reactor, the theoretical model has been
used for implementing a new finite element in a F.E.M. code. This finite element simulates the thermalmechanical behaviour of a ensemble of regular lattices. After some test cases, the new finite element ,
implemented in a FEM code, has been used to simulate an elementary cell of the TBM under operating
loading. The numerical results obtained with the new finite element have been compared with those of the
discrete as well as continuous model, applied at the same geometry and loading. The comparison shows a good
agreement between the discrete and the implemented model in terms of stresses and displacements.
Corresponding Author:
AQUARO DONATO
aquaro@ing.unipi.it
Dipartimento di ingegneria meccanica, nucleare Università di Pisa Via Dio ti salvi, n 1
387
- I - Materials Technology and Breeding Blankets.
P1T-I-269
BEHAVIOUR OF TRITIUM IN BREEDING BLANKET MATERIALS
KIZANE, GUNTA, JURIS TILIKS AIGARS VITINS JURIS TILIKS, JR. JANIS RUDZITIS
Laboratory of Solid State Radiation Chemistry, Department of Chemistry, University of Latvia, 4 Kronvalda Blvd., LV-1010
Riga, Latvia
Li4SiO4 ceramic pebbles as tritium breeder, and beryllium pebbles as neutron multiplier are envisaged for the
ITER and DEMO projects. Tritium forms initially as positively charged ions (T+) as a result of nuclear
reactions in both these materials. Negatively charged (T-), neutral atomic (T0) and molecular (T2 or HT)
tritium can form as a result of electron capture and dimerization. The rate of tritium release is limited mostly
by the volume diffusion in crystal grains, which differs for the different T forms. The diffusion parameters of
the charged T forms can be affected by the magnetic field (MF) up to 6-10 T in the case of HCPB. The
diffusion path lengthens under action of MF (the Larmor effect), which can delay the T release. In order to
predict the T release, the distribution of the T forms in the material has to be known. The abundance ratio of T
forms, its changes during the annealing, and their distribution along a pebble radius were investigated in this
study by dissolving the pebbles under controlled hydrodynamic conditions in the solutions of different
scavengers. The Li4SiO4 pebbles irradiated both weakly (1E18 n/sq.m) and in the EXOTIC-8 experiment
(1E25 n/sq.m), and the Be pebbles from the BERYLLIUM experiment (1E25 n/sq.m) were investigated in this
study. Acid, nitrate and monochloroacetate scavenging systems enabled to determine selectively T+, T0, Tand T2 (HT) forms. The main T form in the weakly irradiated Li4SiO4 pebbles is T+, which determines the
detaining effect of MF in the case of coarse crystal grains. At high values of the fluence, a considerable
amount of T2 (HT) appears as well. Changes in the abundance ratio of T forms take place during the annealing
of ceramic pebbles. The external MF (up to 2.4 T) affects the transformation of these forms, changing the
kinetics of reactions of diffusing T particles with matrix defects. The distribution of the tritium forms along a
pebble radius is not uniform. The most part of T+ is localised in a subsurface layer of thickness to 5E-5 m of a
Li4SiO4 pebble of diameter 0.5 mm. The tritium distribution in the pebble volume becomes more uniform
during the annealing; however this process is affected considerably by the external MF. The obtained results
enable to predict the T release from the pebbles under the real operating conditions of a fusion reactor.
Corresponding Author:
KIZANE, GUNTA
radchem@kfi.lu.lv
Laboratory of Solid State Radiation Chemistry, Department of Chemistry, University of Latvia, 4 Kronvalda
Blvd., LV-1010 Riga, Latvia
388
- I - Materials Technology and Breeding Blankets.
P1T-I-276
MUTUAL CORROSION OF EUROFER97 AND THE BLANKET
CERAMIC MATERIALS
TILIKS, JURIS, GUNTA KIZANE (1) BORIS POLYAKOV (2) AIGARS VITINS (1)
(1) Laboratory of Solid State Radiation Chemistry, Department of Chemistry, University of Latvia, 4 Kronvalda Blvd., LV1010 Riga, Latvia. (2) Institute of Chemical Physics, University of Latvia, 19 Raina Blvd., LV-1586 Riga, Latvia
Mutual compatibility of the structural materials is an important problem of the blanket zone of fusion reactors.
The low-activation martensitic steel EUROFER97 is chosen as the optimal material for HCPB reactor cooling
plates, pipes and casing at present. During the long-term operation (approx. 20,000 hours at 920-1170 K, fast
neutron flux 1E19 n/(sq.m*s)) the physicochemical processes inducing the metal corrosion can take place on
the contact surfaces between metal and ceramics. In addition to these factors at the real operation of the
DEMO or ITER devices, high magnetic fields (MF) up to 10 T will act as well. The MF effect on the mutual
corrosion of metal and ceramics has not been investigated up to now. The effects of high temperature (920 K),
irradiation (5 MeV electrons, the dose rate 15 MGy/h) and MF (1.5 T) were investigated both simultaneously
and separately on the mutual corrosion of polished surfaces of EUROFER97 and ceramic materials (Li4SiO4
and Li2TiO3 powders). The samples were prepared in the "sandwich" type geometry. The changes of the
surface of samples were investigated by atomic force method. Some of the surface characterisation parameters
were determined - roughness (DIN 4718), root mean square (ISO 4287/1) and ten point height (ANSI B 46 1).
The obtained results indicate that the simultaneous action of high temperature (T), radiation (R) and magnetic
field (M) generates the largest changes of the surface. Li4SiO4 caused the greatest surface corrosion of
EUROFER97. The separate action of (T+R), (T+M) and (T) creates a twice-smaller surface roughness.
Radiation damages the metal oxide layer, which protects the surface. MF affects the formation of radiolytic
oxygen, which is an effective corrosion agent.
Corresponding Author:
TILIKS, JURIS
radchem@kfi.lu.lv
Laboratory of Solid State Radiation Chemistry, Department of Chemistry, University of Latvia, 4 Kronvalda
Blvd., LV-1010 Riga, Latvia
389
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P1T-I-287
AUTOMATIC GENERATION OF A JET 3D NEUTRONICS MODEL
FROM CAD GEOMETRY DATA FOR MONTE CARLO
CALCULATIONS
H. TSIGE-TAMIRAT, U. FISCHER(1) P. CARMAN(2) M. LOUGHLIN(2)
(1)Association FZK-Euratom/Forschungszentrum Karlsruhe, Institut für Reaktorsicherheit, Postfach 3640, 76021 Karlsruhe,
Germany (2)Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB, UK
The Monte Carlo technique enables the use of full and detailed 3D geometry models in neutronics calculations.
The manual modelling of a complex geometry with a Monte Carlo code, as it is common practice, is an
extensive and time-consuming task. A more efficient way is to make use of available CAD geometry data in
the Monte Carlo calculations. This can be achieved by converting the CAD data into the semi-algebraic
representation used by Monte Carlo codes such as MCNP. Suitable conversion algorithms have been recently
developed at Forschungszentrum Karlsruhe and were implemented into an interface programme with a
graphical user interface based on CAD kernel and graphics software. Standard CAD interface data files (IGES
and STEP format) can be imported and converted to the MCNP geometry representation. In the frame of the
JET Fusion Technology programme, first successful application tests of the interface programme have been
performed for a full JET octant model generated by CATIA V5 in the STEP format AP-214 at the JET
drawing office. This paper outlines the main features of the interface programme and describes in detail the
processing of the CAD data defining the JET torus sector with all relevant components and the conversion into
an MCNP geometry model. It was shown that the conversion process does not introduce any approximation so
that the resulting MCNP geometry is fully equivalent to the original CAD geometry. The complexity of the
geometry, however, slightly increased in terms of the number of geometry cells and related surfaces. The
converted geometry has been validated by means of subsequent MCNP calculations for the cell volumes.
Corresponding Author:
H. TSIGE-TAMIRAT
tsige@irs.fzk.de
Forschungszentrum Karlsruhe, Institut für Reaktorsicherheit, Postfach 3640, 76021 Karlsruhe, Germany
390
- I - Materials Technology and Breeding Blankets.
P1T-I-289
PERFORMANCE OF A HYDROGEN SENSOR IN PB-16LI
CIAMPICHETTI ANDREA, ITALO RICAPITO (1) GIANLUCA BENAMATI (2) MASSIMO ZUCCHETTI (3)
(1) RSI Sistemi, Milano, Italy (2) ENEA FIS ING, C.R. Brasimone, Camugnano (Bo), Italy (3) EURATOM/ENEA Fusion
Association, Politecnico di Torino, DENER, C.so Duca degli Abruzzi 24, 10129 Torino, Italy
In the HCLL (Helium Cooled Lithium Lead) TBM (Test Blanket Module) for ITER, a correct and reliable
management of tritium is of basic importance, both for safety and fuel cycle reasons. To develop a sensor for
measurements of hydrogen (and its isotopes) concentration in liquid Pb-16Li, a permeating capsule is being
developed. Different simulations with a mathematical model were performed, and then the sensor was
designed, constructed and tested. Niobium was initially chosen as constructing material. The experimental
results, however, both in gas phase and in liquid metal, showed a permeating flux much lower than predicted,
most probably due to the formation of an oxide layer. The sensor could not operate neither in dynamic nor in
static mode in this way. To overcome this problem, another permeable material (Fe Armco) has been chosen,
also taking into account the easy welding and machining of this material. The sensor design has been reviewed
and optimised too, mainly reducing the sensor walls thickness. The new device has been constructed and tested
in the LEDI device in ENEA Brasimone: this device permits a quicker experimental operation than the
previously used device (VIVALDI). The testing temperature was 400-450 C and the external hydrogen
pressure between 250-1100 mbar. Results of testing in gas phase have shown that the sensor cannot be
operated in equilibrium mode yet, at least at the present conditions, because of the very long time necessary to
reach the hydrogen pressure equilibrium. Further optimisation of the sensor geometry could solve the problem
and this will be the next phase of the research, together with the testing of other materials (e.g., composite
materials). Present experimental results are however very positive, since they have shown that the Fe sensor
can be successfully operated in dynamic mode, since it is possible to link, at each temperature, the permeation
fluxes to the hydrogen test pressure by means of a simple power equation: then, hydrogen concentration in the
liquid metal is determined by the Sieverts' law. The Fe-based sensor quickly reaches the steady-state value of
the permeati