Plasma-Surface
Interaction in the
Be/W Environment:
from the JET-ILW
to ITER
S. Brezinsek and JET-EFDA contributors
S. Brezinsek 27.05.2014
/ 21st PSI Kanazawa
21st PSI conference, Kanazawa,
/ 27.05. 2014 / 1(37)
1
Armour Materials for Fusion Devices
Components Lifetime
Plasma Performance
Tokamak Safety
0
10
Dilution
-1
10
Be
-2
10
10-3
W
BeC
BeO+C
WC
-4
10
0
Low-Z: strong wall erosion
100
Tw JET
200
300
400
500
600
temperature [°C]
Co-deposition dominates
long-term retention in C
High-Z: low sputtering /
mainly by impurities
Metals: potential melting
C
CPS 13.489-6
Radiation
t ritium concentration [T/X]
PISCES
Maximum allowed reactor
concentration for W: ~10-5
Metals: low retention
ITER / DEMO / Reactor: Neutrons and neutron damage need to be considered
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 2(37)
2
ITER: Plasma-Facing Material Selection
ITER
 ITER PFM selection (till 10/2013)
 Be/W/C in the non-active phase (H/He)
 Be/W in the active phase (DD/DT)
 Need for an integrated tokamak experiment in
the ITER-material mix identified in the community
 Decision to transform JET from an all-C PFC
device (JET-C) into an ITER “test bed “ (~2006)
 Option A: Be/W/C
 Option B: Be/W
Beryllium
ITER
 Option B selected in (2008): (i) to give timely input
for DT and (ii) to start with “clean” conditions
Tungsten
 JET-ILW: Be first wall and W divertor
http://www.iter.org
Carbon
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 3(37)
3
ITER: Plasma-Facing Material Selection
ITER
ITER
 ITER PFM selection (till 10/2013)
 Be/W/C in the non-active phase (H/He)
 Be/W in the active phase (DD/DT)
 Need for an integrated tokamak experiment in
the ITER-material mix identified in the community
 Decision to transform JET from an all-C PFC
device (JET-C) into an ITER “test bed “ (~2006)
 Option A: Be/W/C
 Option B: Be/W
Beryllium
Beryllium
ITER
 Option B selected in (2008): (i) to give timely input
for DT and (ii) to start with “clean” conditions
Tungsten
Tungsten
 JET-ILW: Be first wall and W divertor
http://www.iter.org
http://www.iter.org
Tungsten
Carbon
 ITER PFM selection revised in 10/2013 owing to
new experimental results from tokamaks, W PFC
development, extensive ITPA review, and cost
reduction: Be/W in all ITER operational phases
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 4(37)
4
Road to Metallic Plasma-Facing
Components in Fusion Devices
 Metallic wall studies in 1970s (e.g. W in PLT)
=> unfavourable transport / W accumulation
 Use of graphite PFCs and progress in plasma
performance, but too high fuel retention
 Renaissance of metallic walls (reactor needs):
=> low erosion and long lifetime
=> low retention and sustainable tritium cycle
W
 Set of devices with full or partial metallic PFCs:
AUG (all-W - 2007), CMOD (Mo), FTU (Mo), JET
(Be/W - 2011), TEXTOR (limiter PFCs) …
 Similar observations in all metallic devices
used to increase physics understanding
ASDEX Upgrade
Mo
G.F. Matthews JNM 2013
CMOD
Here: focus on JET- ILW and Be/W exploitation
 Demonstrate low fuel retention, migration
and possible fuel recovery
 Demonstrate plasma compatibility in Be/W
 Gain physics understanding in Be/W plasmas
 Secure safe ITER operation and exploitation
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 5(37)
5
Road to Metallic Plasma-Facing
Components in Fusion Devices
 Metallic wall studies in 1970s (e.g. W in PLT)
=> unfavourable transport / W accumulation
 Use of graphite PFCs and progress in plasma
performance, but too high fuel retention
 Renaissance of metallic walls (reactor needs):
=> low erosion and long lifetime
=> low retention and sustainable tritium cycle
W
ASDEX Upgrade
Mo
 Set of devices with full or partial metallic PFCs:
AUG (all-W - 2007), CMOD (Mo), FTU (Mo), JET
(Be/W - 2011), TEXTOR (limiter PFCs) …
 Similar observations in all metallic devices
used to increase physics understanding
CMOD
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 6(37)
6
Impact of the ILW on PSI / Plasma
III: Change in
fuel retention
and recycling
II: Change in
erosion and
material migration
I: C replaced by Be
and high plasma purity
VI: Restricted operational
window with W divertor
V: Change in
plasma conditions
and confinement
IV: Change in
impurity content
and radiation
 More changes with ILW introduction than expected and predicted by EDGE/SOL/PSI modelling
 In JET-C some of the EDGE/SOL/PSI physics was masked and overlaid by the impact of C on
the plasma and revision of the role of some processes is required and currently ongoing
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 7(37)
7
Outline
 I: ITER-Like Wall
 PFCs and design
 Power handling Be/W
 Residual C levels
 II: Impurity sources and migration
 III: Fuel recycling and retention
 IV: Radiation and impurity concentrations
 V: Plasma behaviour and confinement
 VI: Operation with W divertor
 Summary
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 8(37)
8
Plasma-Facing Components
in the JET ITER-Like Wall
 Remote handling installation of all PFCs: a logistical/technical example for ITER RH
Additional W-coated CFC
protection tiles installed
in the main chamber
 More than 80 000 part with 350 tools installed by RH
 All PFCs are only inertially cooled by which is different to actively cooled ITER PFCs
 PFCs are optimised designed with respect to power handling and material stress [V. Riccardo Phys Scr. 2008]
 Power handling predictions verified by experiments for Be limiters and W divertor [G.Arnoux Phys. Scri.. 2014]
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 9(37)
9
Be PFC Power Handling:
Power-Decay Length at Tile Centre
 Dedicated Be PFC power handling experiments to quantify the power decay length
in limiter configuration: support of ITER PFC design qualification process
PCFFLUX
G. Arnoux Phys. Scr. 2014
 Predicted power handling and applied code [PFCFLUX] verified by dedicated experiments, but
unexpected narrow power decay length at the limiter centre with enhanced power load
 Minor toroidal asymmetry (~1mm) lead to local Be melting, but didn’t restrict normal operation
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 10(37)
10
Bulk W PFC Melting Experiment
 Dedicated W melting experiment in JET carried to verify MEMOS code used for ITER
prediction to access the impact of full W divertor damage on plasma and operation
 High additional local power load due to installed mismatched lamella (1GW/m2 per ELM)
 Shallow melting by ELM impact on top of high steady-state thermal load
 No disruptions caused by melting or droplets
 MEMOS benchmark progressing
 Post-mortem analysis pending
[J.W. Coenen I8]
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 11(37)
11
Carbon Wall (JET-C) vs.
Beryllium/Tungsten Wall (JET-ILW)
 Exchange of PFC without active cleaning. Test of wall conditioning techniques.
JET-C
JET-ILW
JET
2009
CFC
S. Brezinsek
JNM 2011
G.F. Matthews
Phys. Scr. 2012
 Full metallic wall installation at once – in contrast to step-wise approach of AUG [R. Neu JNM 2013]
 No active cleaning of vessel interior before new PFC installation
 Only baking (up to 320°C) and D-GDC (150h) applied for conditioning purpose prior first plasma
 No inter-shot D-GDC, no Be evaporations applied for operational purpose [D. Douai JNM 2013]
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 12(37)
12
Suppression of Residual Carbon
with ILW
 Very low residual C content in the plasma – high plasma purity
 ILW is a good approximation of conditions to be expected in ITER starting with Be/W
 C dropped with ILW installation by one order of magnitude (edge and core)
 Be is gettering O via BeO production, but still O present at low levels in plasma
5
I (OVI at 103.2nm) / ne [arb. units]
JET with W HD divertor
and Be first wall
deuterium plasma
0
10
-1
10
-2
10
x20
GB divertor
I (CIII at 97.7nm) / ne [arb. units]
Helium
HD
JET with CFC
HD divertor
deuterium plasma
C28-C30
C31
-3
10
70000
10
JET with CFC
HD divertor
deuterium plasma
4
10
Helium
HD
JET with W HD divertor
and Be first wall
deuterium plasma
3
10
2
10
1
10
0
10
GB divertor
1
10
C28-C30
C31
-1
72000
74000
76000
78000
80000
82000
84000
86000
10
70000
72000
74000
76000
JET pulse number
78000
80000
82000
84000
86000
JET pulse number
S. Brezinsek JNM2013, J.W. Coenen NF2013
 Averaged Zeff dropped from 2.0 (JET-C) to 1.2 (JET-ILW)
 If primary source strength would be constant: Zeff of 1.6 would have been expected
 Changes comparable to He operation in JET-C (absence of chem. sputtering of C in He)
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 13(37)
13
Outline
WI at 400.9nm
#JPN 81182
inner
divertor
 I: ITER-Like Wall
 II: Impurity sources and migration
 W sources
 Be sources
 Migration pattern
 Deposition pattern
III: Fuel recycling and retention
 IV: Radiation and impurity concentrations
 V: Plasma behaviour and confinement
 VI: Operation with W divertor
 Summary
outer
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 14(37)
14
Effective W Sputtering Yields
 W erosion yield governed by Be impurity ion bombardment - according to TRIM
 W source strength is lower than in AUG if compared at the same impact energy (or Te)
Inter-ELM and steady-state regimes with
low Te show low W sputtering yields
1.E-01
1.E-02
5.7% C4+
4.0% C4+
2.0% C4+
1.0% C4+
0.5% Be 4+
0.5% Be 2+
W/ion
1.E-03
1.E-04
ASDEX Upgrade
1.E-05
JET, density steps
1.E-06
JET, density ramp
TEXTOR
1.E-07
0
20
40
60
Te [eV]
80
100
R. Dux JNM 2009
G.v. Rooij JNM 2013
 Plasma cooling by N2 / Ar seeding till Eimp drops below sputtering threshold - as in AUG [R.Neu JNM 2011]
 Residual W source in seeded discharges is determined by ELMs - as in AUG [R.Dux JNM2009]
 Prompt re-deposition at least 50%. PIC modelling predicts higher values [D.Tschakaya this conference]
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 15(37)
15
Effective Be Sputtering Yields
 JET-ILW experiments are used to the verify ERO code and involved atomic data
applied for lifetime predictions of ITER blanket modules: revision required
divertor
plasmas
limiter
plasmas
D. Borodin
Phys. Scri. 2014
Db 486nm
Be II at 527nm
 Inner-wall limited plasmas executed at large range of fuelling rates: Te and ne scan
 Dominant self-sputtering observed at high electron temperatures / impact energies
 Comparison with Be sputtering yields from initial ERO run (phys. sputtering) in fair agreement
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 16(37)
16
Chemical Assisted Physical Sputtering
of Be via BeDx
 Chemical assisted physical sputtering of Be via BeD needs to be considered
2.0
#82626
at 11s
# 82626
Tsurf
spot
size
Height[m]
1.0
0.0
-1.0
inner wall
Be limiter
KS3 view
HFS
2.0
LFS
2.5
3.0
3.5
Radius [m]
𝑝ℎ𝑦𝑠
𝑡𝑜𝑡
𝑌𝐵𝑒
(𝐸𝑖𝑛 , 𝑇𝑠𝑢𝑟𝑓 , 𝜃) = 𝑌𝐵𝑒
4.0
𝑐ℎ𝑒𝑚
+ 𝑌𝐵𝑒
75% 𝑣𝑖𝑎 𝐵𝑒𝐷 + 𝑒 → 𝐵𝑒 + 𝐷 + 𝑒′
25% 𝑣𝑖𝑎 𝐵𝑒𝐷 + 𝑒 → 𝐵𝑒𝐷 + + 𝑒′′ + 𝑒′
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 17(37)
S.Brezinsek
EPS 2013
17
Chemical Assisted Physical Sputtering
of Be via BeDx
 Chemical assisted physical sputtering of Be via BeD needs to be considered
 Confirms Molecular Dynamics predictions
at impact energies of 75eV [C. Björkas NJP 2009]
 Explains partially variety of Be sputtering yields
observed e.g. in PISCES [D. Nishijima PPFC 2008]
 Energetic threshold at low impact energies –
not like carbon chemistry!
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 18(37)
S.Brezinsek
EPS 2013
18
Material Migration: Main Chamber
SOL flows
HFS
Ions
Ions
CXN
CXN
Limiter configuration
 Sputtering at poloidal limiters at high Te / impact
energies (Ein≥75eV)
 At high Ein: Be sputtering yield (JET-ILW) is larger
than C sputtering yield (JET-C)
 Campaign averaged erosion rate at centre tile:
Spectroscopy : 4.1x1018 Be/s gross erosion
Post-mortem analysis: 2.3x1018 Be/s net erosion
 Moderate increase of total limiter source (25%)
 Majority of eroded Be remains in the main
chamber and is redistributed on limiter/wall
 10Be tracer confirms redistribution
LFS
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 19(37)
A: Widdowson Phys. Scr. 2013
S. Brezinsek ICFRM2013
I. Bykov this conference
19
Material Migration: Main Chamber
SOL flows
HFS
Ions
Ions
CXN
CXN
Divertor configuration
 Sputtering at inner wall and limiter at low Te /
impact energies (Ein<10eV)
 At low Ein: Be sputtering yield (JET-ILW) is lower
than total C sputtering yield (JET-C)
 Spectroscopy and post-mortem analysis revealed
a factor 4-5 smaller primary source with ILW (Be in
JET-ILW vs. C in JET-C)
 Absence of chemical erosion by low energetic
particles in case of ILW!
 Majority of residual Be transported in SOL towards
inner divertor. Minor fraction reaches outer leg.
LFS
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 20(37)
S. Krat sub. JNM 2013
S. Brezinsek IAEA2014
P. Petersson this conference
20
Be Migration within Divertor
 Role of C chemistry previously underestimated: Material migration to remote areas is
lower with Be. Consequences e.g. for fuel retention and removal techniques in ITER
 Integral deposition in divertor is low: factor ~7 less when compared to JET-C (time normalised)
ERO
1
3
5
4
[A. Kirschner I15]





Deposition monitor: H.G. Esser, J. Beal this conference
Arriving Be deposited primarily on the apron of inner divertor PFCs
Suppressed multistep Be transport and not full Be coverage of vertical target
Be only physically re-eroded at strike-point and transported towards LFS (factor ~15 less)
Suppressed Be transport to remote areas (factor ~30-50) / less configuration changes
Residual C undergoes multi-step transport enriches at remote areas! C, N, O present
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 21(37)
21
Be Migration within Divertor
 Role of C chemistry previously underestimated: Material migration to remote areas is
lower with Be. Consequences e.g. for fuel retention and removal techniques in ITER
 Integral deposition in divertor is low: factor ~7 less when compared to JET-C (time normalised)
ERO
1
3
5
4
[A. Kirschner I15]





Deposition monitor: H.G. Esser, J. Beal this conference
Arriving Be deposited primarily on the apron of inner divertor PFCs
Suppressed multistep Be transport and not full Be coverage of vertical target
Be only physically re-eroded at strike-point and transported towards LFS (factor ~15 less)
Suppressed Be transport to remote areas (factor ~30-50) / less configuration changes
Residual C undergoes multi-step transport enriches at remote areas! C, N, O present
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 22(37)
22
Material Migration Balance:
JET-ILW vs. JET-C
 Material erosion reduced with ILW. Positive impact on PFC lifetime and ITER safety.
JET-ILW (2011-2)
(6h lim / 12h div)
JET-C (2007-9)
(12 lim /21h div)
~30g (-)
~273g (-)
< 2g(±)/melting
130g (-)
0
0.8g (+)
IWGL centre (1 row)
8g (-)
11g (-)
IWGL bottom (1 row)
0
No measured
15g (-)
129g (-)
0
No measured
OPL centre (1 row)
5g (-)
3.1g (-)
OPL bottom (1 row)
0
No measured
~ 37 g (+)
~459g (+)
HFGC
10g (+)
30g (+)
Tile 1
25g (+)
65g (+)
Tile 3,4,6
<6g (+)
428g (+)
Tile 7,8
<4g (-)
64g (-)
Dust :
1g
233 g
Main chamber (SUM):
Dump Plates
IWGL top (1 row)
Inner Wall Cladding (all)
OPL top (1 row)
Divertor (SUM):
A. Baron-Wiechec this conference
Main difference between JET-C and JET-ILW: Absence of chemical erosion (by ions
and neutrals) at low impact energies and, thus, Te in the case of Be and JET ILW.
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 23(37)
23
Outline
 I: ITER-Like Wall
 II: Impurity sources and migration
 III: Fuel recycling and retention
 Long-term fuel retention and outgassing
 Isotope exchange and fuel removal
 Outgassing during ELMs
 IV: Radiation and impurity concentrations
 V: Plasma behaviour and confinement
 VI: Operation with W divertor
 Summary
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 24(37)
24
Long-Term Fuel Retention
 Gas Balances used for ITER tritium retention predictions and input to WallDYN
V. Philipps
PFMC 2013
S. Brezinsek
NF2013
 Reduction of fuel retention rate by more than one order magnitude [confirmed by PMA: Heinola O10]
 Long-term retention mechanism: implantation and co-deposition (dominant)
 Reasons for the reduction from JET-C to JET-ILW:
 Be primary source and Be transport to divertor smaller than C in JET-C
 Lower fuel content in pure Be co-deposits in comparison with C co-deposits
 Integral outgassing comparable between JET-C and JET-ILW
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 25(37)
25
ITER Fuel Retention Predictions
 Validated WallDYN code predicts operational time before ITER tritium limit is reached
[K. Schmid I6]
 WallDYN predicts retention ratio of C to Be/W of 10 to 100
 WallDYN predicts flux scaling of 〖~Φ〗^0.6
 ITER „plasma scan“ show large scatter (factor 10)
Fuel retention limit:
 ITER with pure C walls:
100-700 discharges
 ITER with Be+W walls:
3000-20000 discharges
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 26(37)
[full DT 400s plasmas]
26
Isotope Exchange with JET-ILW
 Assessment of plasma isotope exchange discharges for fuel replacement (cleaning)
T. Loarer this conference
 Faster cleaner plasmas with JET-ILW than in JET-C: 95.5% D2 reached within 150s
 Accessible gas inventory by plasma amounts ~2-3x1022 in comparison with 2x1023 in JET-C
 Fraction of released H is co-deposited in the discharge again rather than pumped away. ICWC
more effective and releases twice the fuel measured by plasma exchange [T.Wauters this conference]
 Disadvantage: gas needs to be refilled in next discharge to obtain request isotope ratio
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 27(37)
27
Low Recycling Conditions of W PFCs
 Low fuel inventory in W and its outgassing capabilities impact on ELM cycle
R<<1
R=1
V. Philipps PFMC2013
Low ne,ped
W divertor phase
High ne,ped
Be limiter phase
B. Sieglin PPCF2013
L-mode #81938-73
 W divertor PFCs need to be filled
with plasma fuel before R=1 reached
Pedestal crash
Pedestal fuelling prolonged
D refuelling via
recycling (R<1)
 Longer ELM duration with ILW (MHD same)
 JET-C and JET-ILW similar at high Tped
Particles and heat stream to target
W target plate is heated up
Deuterium reservoir
partially emptied
D is desorbed
First desorption
peak on W-coated CFC
at 300°C
 Reservoir of fuel in C was practically unlimited due to the large retention in co-deposits
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 28(37)
28
Outline
 I: ITER-Like Wall
 II: Impurity sources and migration
 III: Fuel recycling and retention
 IV: Radiation and impurity concentrations
 Zeff and impurities
 Divertor operation and modelling
 Disruptions and runaway electrons
 V: Plasma behaviour and confinement
 VI: Operation with W divertor
 Summary
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 29(37)
29
Impurity Concentration / Zeff
 Zeff in JET-ILW lower than expected. Impact on Zeff in ITER simulations to be checked
Zeff fro m C XRS
A. Kallenbach NF 2009
 JET-ILW shows an averaged Zeff in H-mode of about 1.2
without any additional conditioning (Zeff~2 in JET-C)
 Averaged concentrations: CC~0.1% and CBe~1.0%
 AUG (~1.7) and CMOD (~1.8) show higher Zeff in normal
operation in H-mode without wall conditioning
 Even with boronisation (suppression of C, O impurities) the
averaged Zeff in AUG falls not below ~1.4
2.0
1.8
1.6
JET-C no N2
JET-ILW with N2
1.4
1.2
JET-ILW no N2 :av. Zeff from Bremsstrahlung
3.2
3.4
major radius [m]
3.6
3.8
C. Giroud EPS 2014
 With nitrogen seeding: JET-ILW returns to averaged Zeff
values about 1.5-1.7 and profiles close to JET-C (w/o N2)
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 30(37)
30
Edge Modelling of
JET-ILW L- and H-mode Plasmas
 Benchmark of EDGE2D-EIRENE and SOLPS in D and N2-seeded plasmas
 L- and H-mode plasmas in D revealed a higher density limit due to loss of C radiation
 EDGE2D-EIRENE and SOLPS can partially reproduce behaviour [M.Groth NF2013 / C.Guillemaut sub. NF /
K.Lawson this conference / L. Aho-Mantial JNM2013]
 H-mode plasmas with N2 seeding achieved full detachment at outer target plate [C. Giroud NF2013]
 EDGE2D-EIRENE reproduces all phases from low recycling to full particle detachment at LFS
 Sequence of power detachment (nitrogen radiation), momentum detachment (D ion-neutral
friction) and particle detachment (D volume recombination)
[A. Jarvinen I18]
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 31(37)
31
Runaway Electron Generation
with the JET-ILW
 Verify physics understanding and prediction of runaway electron (RE) formation in ITER
 RE generation without massive gas injection not present in JET-ILW in contrast to JET-C
Growth rate for RE could be governed by Zeff [Y. Igitkhanov]
 Active RE generation by high fraction massive Ar injection into JET-ILW plasma
JET-C
JET-ILW
[C. Reux I19]
 Similar boundary for Ar-induced runaways in JET-C and JET-ILW
 RE production rate is higher in JET-ILW in comparison with JET-C
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 32(37)
32
Outline
 I: ITER-Like Wall
 II: Impurity sources and migration
 III: Fuel recycling and retention
 IV: Radiation and impurity concentrations
 V: Plasma behaviour and confinement
 L-H transition
 H-mode confinement and pedestal
 ELMs
 VI: Operation with W divertor
 Summary
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 33(37)
33
L-H Transition: JET-C vs. JET-ILW
 Favourable minimum in L-H transition observed in the high density branch in D
 Pure D plasma show reduction of threshold power as function of magnetic field, density
and configuration at typical low Zeff for JET-ILW [C.F. Maggi NF2014]
 Strong nitrogen seeding and associated increase of Zeff recover almost JET-C behaviour
 Two mechanisms are being explored:
 The effect of low-Z on the stability of the background edge turbulence [C. Bourdelle NFL 2014]
 Changes in divertor and SOL radiation patterns between JET-C and JET-ILW
C.F. Maggi EPS2014
H-L back transition: A. Huber O17
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 34(37)
34
H-mode Confinement
 ITER baseline scenario based on JET-C like high triangularity plasmas - recovery?





JET-ILW and AUG show degradation of confinement with respect to carbon wall operation
Degradation partially governed by fuelling requirements to allow safe operation in W
Degradation is not in the plasma core, but determined by changes in the pedestal
Strongest degradation for JET high triangularity plasmas by about 25%
Pedestal recovery possible by increase of bN or by impurity seeding with med. Z [ G. Giroud NF2013]
[R.Neu JNM2013]
M. Beurskens and J. Schweinzer
NF2014
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 35(37)
35
H-mode Pedestal
 ITER baseline scenario based on JET-C like high triangularity plasmas
JET
Te( ped ) (k eV)
 Identification of the cause of
of pedestal degradation is
currently highest priority
 Different models proposed, but
not yet able to reproduce the
observations in JET-C AND
JET-ILW
1.2
related to the observed change
in recycling combined with a
different Zeff profile in the edge
1.0
0.8
0.6
0.4
0.2
0.0
0.0
hig h d,
JET-C
low d, JET-C
hig h d, JET-ILW
low d, JET-ILW
0.5
1.0 1.5
<Zeff>
7
5
wit h N2
wit h N2
2
4
6
8
ne(ped) (x1019 m -3)
10
ne( pe d) [1019 m-3]
 Alternative proposals are
9
hig h d,
JET-C
low d, JET-C
hig h d, JET-ILW
low d, JET-ILW
0.5
0
Te(p ed) [k eV]
[Beurskens/Schweinzer NF2014]
11
0.0
 Change in pedestal width and in
stability is a strong candidate
1.0
wi th N2
wi th N2
2.0
2.5
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 36(37)
3 kPa
M. Beurskens NF2014
10
8
6
4
2
0
0.0
hig h d,
JET-C
low d, JET-C
hig h d, JET-ILW
low d, JET-ILW
0.5
1.0 1.5
<Zeff>
wi th N2
wi th N2
2.0
2.5
36
Outline
NBI (19MW)
NBI (16MW) +
RF (3MW)
 I: ITER-Like Wall
 II: Impurity sources and migration
 III: Fuel recycling and retention
 IV: Radiation and impurity concentrations
 V: Plasma behaviour and confinement
 VI: Operation with W divertor
 W control mechanisms
 Steady-state conditions
 Summary
W in SXR range
W in SXR range
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 37(37)
37
Stable H-mode Operation
 Operational window narrower with JET-ILW: unfuelled plasmas are unstable
 Limited access to low or no gas fuelling which provided best performed in JET-C
 Stable type I ELMy H-modes
B=2.0T,
I p=2.0MA, Z eff=1.2
t
JPN 85290
MW
D
MJ
keV
20
10 m
-3
Pin
arb. u.
15
10
5
0.0
10.0
6.0
Prad
in low and high triangular plasmas
 Stable operation requires either
 Gas fuelling (to reduce W source)
 Minimum ELM frequency (to flush W)
2.0
 Central heating (to avoid W peaking)
12
10
8
6
5
4
3
2
1
4.0
3.0
2.0
1.0
1.0
0.8
0.6
ne(0)
otherwise W-accumulation might occur
 W concentration below CW~5x10-5
Te(0)
Th. Puetterich PPCF2013 [N. Fedorczak I9]
8
WMHD
Example:
H98y
 Stable ELMy H-mode with H98~1.0 (Ip=2.0
MA, Bt=2.0 T) and fuelling achieved
10
12
14
time [s]
16
18
20
 Operation with strike-points close to
E. de la Luna
et al.
pump duct entrance [P. Tamain O33]
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 38(37)
38
W-compatible Integrated Scenario
 ITER requires a stationary integrated scenario: core and edge compatible
 Qualification of nitrogen as suitable gas for ITER seeding
MW
JET –ILW example with N2:
22
1020m-2 10 s
-1
Input power:
 15MW NBI for 15s
 3MW RF for 13 s
MJ
keV
Confinement:
 H98~ 0.85
 fGW~ 0.8
 Zeff~ 1.6
MHD
time [s]
C. Grioud EPS 2014
 Confinement increase by 40%
with respect to unseeded case
 Semi-detached divertor
operation in both legs
 Stationary N seeded scenario
achieved
 No feed-back required
 Te below sputtering threshold of
W by N
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 39(37)
39
Summary
 I: Be replaced C as main impurity in JET: ideal test bed to study PSI like in ITER
 II: Reduction of material erosion and migration demonstrated in JET-ILW
 III: Reduction of fuel retention and recycling demonstrated in JET-ILW
IV: High purity and low impurity content and radiation demonstrated in JET-ILW
 V: A part of the EDGE/SOL/PSI physics was masked and overlaid by the impact of C
on the plasma and revision of the role of some processes is ongoing. As main
driver for the changes recycling/retention and impurity content/Zeff are discussed
VI: Operational space with W divertor in the JET-ILW partially explored and lower
boundaries found: Tools for W control are developed and are at hand.
JET equipped with Be/W PFCs is providing strong input to ITER in the area of PSI
which includes design verification, physics understanding and code validation!
 ITER must operate with impurity seeding to mitigate heat loads (and W sputtering)
 ITER will operate with high density divertor and at medium Zeff of 1.7
 Impurity seeding in JET-ILW with medium Z recovers a large fraction of conditions
found in JET-C. Reinterpretation of JET-C is required to validate physics models.
JET-ILW will also in future support ITER to secure safe operation and exploitation
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 40(37)
40
JET-ILW Contributions to this PSI
I: ITER-Like Wall
J.W. Coenen I8, D. Douai I20, G. Arnoux, B. Bazylev, D. Frigione
II: Impurity sources and migration
A. Kirschner I15, M. Arilia, A. Baron-Wiechec, J. Beal, D. Borodin, I. Bykov,
H.G. Esser, N. den Harder, J. Karhunen, P. Petersson, D. Tshakaya
,
III: Fuel recycling and retention
K. Schmid I6, K. Heinola O10, H. Bergsaker, A. Drenik, T. Loarer, T. Wauters
IV: Radiation pattern and impurity concentrations
C. Reux I19, L. Aho-Mantila O27, A. Huber O17, S. Potzel 035, M. Groth,
Y. Igitkhanov, K. Lawson
V: Plasma behaviour and confinement
A. Scarabosio I1, D. Carralero I16, A. Jarvinen I18, P. Tamain O33, D. Harting,
R. Zagorski, G. Telesca
VI: Operation with W divertor
N. Fedorczak I9, M. Sertoli O15, R. Colas O31, E. Lerche
My thank goes to the JET team of the last 4 years which helped to bring the
machine in operation and to explore the interesting PSI and edge physics!
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 41(37)
41
Reserve Slides
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 42(37)
42
Outlook
The JET-ILW represents an ideal test bed for the ITER material mix studies







Ar, Ne seeding as alternative seeding gases for ITER
H and He operation to provide input for initial ITER phase
Impact of He on W an Be PFCs (blisters/fuzz etc.)
Second fast valve for combined disruption mitigation studies
Fuel removal and cleaning techniques comparison
Pellet pacing studies with improved injection line
Extensive edge modelling activities with multiple ITER-relevant code tests
 DT preparation and T, DT campaign
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 43(37)
43
ELM duration
For the same pedestal pressure
ELM duration via
Heat load to the target
Short duration: High Te /low ne
Long duration: Low Te /high ne
B. Sieglin EPS 2013 / PPCF 2013
Change in recycling at target plate
required in order to reproduce
long ELM duration
[D. Harting this conference]
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 44(37)
44
Confinement Recovery with N2 Seeding
at Low and High Triangularity
 Confinement Recovery in Impurity Seeded Discharges in Be/W environment
2.5MA/2.7T
20.
PNBI
10.
4.
2.
PRF
12.
D
4.
6.
N
2.
5.0
3.0
Wmhd(MJ)
1.8
1.4
1.0
bN
1.0
0.8
H98
0.35
0.20
d
10
C. Giroud et al.
12
14
Time (s)
16
18
20
Input power:
 16MW NBI for min. 10s
 3MW RF
Confinement:
#85415 : D2 only
=> Typ. ILW confinement
#85417: N2 and D2 – low d
=> Confinement: +15%
#85419: N2 and D2 – high d
=> Confinement: +40%
 High triangularity shows
best confinement as in
JET-C (at high density)
 Semi-detached divertor
operation in both legs
with seeding
 Similar to AUG [R.Neu JNM2013]
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 45(37)
45
How is all interconnected?
Requires
fuelling or
seeding
Requires
disruption
mitigation
Higher
density
limit
W sputtering
possible
„slower“
disruptions
stable
MARFE
„hotter“
divertor
less radiated
energy
in disruptions
Reduction in radiation
(i.e. in the divertor)
FOR
L-MODE
STUDIES
absence
of RE beams
Requires
ELM control
Requires power
& energy control
Requires
W control
Large W sputtering
due to ELMs
risk of W
melting
risk of W
accumulation
impurity seeding
with moderate N2
more input
power
low fuelling /
optimise shape
Reduction in C content
Confinment recovery
Increase in Be content
Change in fuel
inventory
Change
in recycling
Impurity seeding
or area spreading
Low long term
fuel retention
Higher dynamic
retention
FOR
baseline
H-MODE
STUDIES
Lower confinement at same Paux & Gas
Weaker pedestal pressure
„colder“ pedestal
lower L-H
threshold
Less fuel content
in layers/
low implantation
larger outgassing
stable breakdown
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 46(37)
change in ELM
characteristic
S. Brezinsek
EPS 2012
46
W Transport and Accumulation
 Central heating (ICRF) can be used to heat up core and enhance transport to remove
W from the central region and supress neo-classical accumulation
Density of Prad by W in SXR range [kW/m3] :
NBI (19MW)
NBI+ICRF ([13 +6]MW)C
25
18.8
12.5
6.25
0
M. Goniche
EPS2014
V. Bobkov
et al.
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 47(37)
47
W Control in Seeded Discharges
- Role of ICRH
 In the JET-ILW we see at above 3MW ICRF sufficient central heating to remove
W from the inner core region (similar to AUG with ECRH)
 Impurity control was tested in nitrogen seeded plasmas with and without ICRH
under otherwise identical plasma setup ( VT in high d with D=N=2.2x1022 el/s)
 Impurity control and stationarity achieved
NO ICRH
WITH ICRH
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 48(37)
48
W Control – ELMs and Gas Fuelling
 Minimum gas fuelling required to keep inter-ELM source low (ELM frequency high)
 Minimum ELM frequency required to remove W from pedestal region
 Without minimum gas fuelling no possibility to avoid W accumulation (steady cond.)
 JET-ILW has a “MAIN CHAMBER” W source at an unidentified position
 Be evaporation covered W and Ni (from inconel) and reduced W for a short period
 Limiter discharges ALSO showed W source
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 49(37)
49
Reliable Breakdown with JET-ILW
P. De Vries et al.
Nuclear Fusion 2013
 No issues with non-sustained breakdowns
 C radiation much reduced during breakdown with respect to JET-CFC
 Strong outgassing of fuel between discharges ensures better control with Be walls
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 50(37)
50
Residual Oxygen Content–
Gettering Properties of Be
Oxygen concentration in the plasma edge layer
 Oxygen leak at the start of operation
 Oxygen content below the best levels in well-conditioned JET-C device
 No need for glow discharges for active cleaning of first wall
D. Doaui et al. JNM 2013
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 51(37)
51
Higher Density Limit in JET-ILW
L-mode Plasmas
 Input to benchmark of EDGE2D-EIRENE and SOLPS in D plasmas




Higher density limit in JET-ILW case with complete ion flux roll-over at LFS occurs
Higher gas throughput and neutral pressure at ion flux roll over
Radiation dominated by deuterium in L-mode plasmas
20% less radiation in case of JET-ILW: difference Be and C radiation at low Te
M. Groth NF 2013
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 52(37)
52
Disruptions: JET-ILW vs. JET-C
 Absence of C as main radiator causes strong changes in disruption behaviour
Disruption characteristics with
the ILW in comparison to JET-C:





Lower radiated power
Slower current quench
Higher wall heat load
Longer halo current
No large runaway electron flux
Need for disruption mitigation
at plasma currents > 2.5MA
M. Lehnen JNM 2013
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 53(37)
53
ITER divertor
ITER-like wall in JET
(actively cooled)
(inertial cooling)
melting
recrystallization
surface temp.
surface temp.
Plasma-Facing Components
in ITER and JET
brittle
0
time
450 s
Tmelt
DBTT
0
10 s
S. Brezinsek / 21st PSI Kanazawa / 27.05. 2014 / 54(37)
time
J. Linke et al.
54