IAEA Technical Meeting on on Re‐evaluation of Maximum Operating Temperatures and Accident Conditions for High Temperature Reactor Fuel and Structural Materials, 10‐12 July 2012, Vienna, Austria
Re‐evaluation of Maximum Fuel Temperature of the HTTR at Normal Operation
Hirofumi OHASHI
Nuclear Hydrogen and Heat Application Research Center
Japan Atomic Energy Agency (JAEA) Outline
1. HTTR overview
2. Evaluation of HTTR fuel temperature at design stage
 Evaluation method
 Evaluation results
3. Re‐evaluation of fuel temperature
 1st modification using the operation data of the rise‐
to‐power test up to 850oC
 2nd modification using new analysis model
 3rd modification using the 950˚C operation data
4. Related future tests using HTTR 5. Summary
1
High Temperature Engineering Test Reactor (HTTR) Major specification
HTTR
Fuel Rods
Intermediate heat
exchanger
(IHX)
Containment vessel
Graphite Block
Reactor pressure vessel
Hot‐ gas duct
Fuel
Uranium enrichment
Core
Fuel assembly
Moderator
Primary coolant
Thermal power
Inlet temperature
Outlet temperature
Primary coolant pressure
Primary coolant flow rate




Low enriched UO2
3~10wt% (avg. 6wt%)
Prismatic
Pin‐in‐block
Graphite
Helium
30 MW
395C
850oC / 950C (Max.)
4 MPa
12.4 / 10.2 kg/s
First criticality : 1998
Full power operation (850oC/30 MWt): 2001
950oC operation at full power: 2004
50‐days continuous operation at high outlet temperature (950oC/30MWt) : 2010
2
Structure of Fuel Assembly
Fuel handing hole
Fuel kernel,600μm
High density PyC
SiC
Low density PyC
920μm
Dowel pin
Plug
Fuel compact
Graphite sleeve
8mm
580mm
Coated fuel particle
39mm
34mm
26mm
Fuel compact
Fuel rod
Dowel socket
360mm
Fuel assembly
3
Reactor Core Structure
Control rod
standpipe
Reactor pressure
vessel
Permanent
reflector
Replaceable
reflector
Fuel assembly
Core restraint
mechanism
Reactor pressure
vessel
Core restraint
mechanism
Permanent
reflector
Replaceable
reflector
Core support
plate
Control rod
guide block
Primary helium
gas tube
Fuel assembly
4
Evaluation Method of HTTR Fuel Temperature
(1)
Nuclear design code
Power density and neutron fluence distributions
Fuel, control rod, core component, core internal structure design data
FLOWNET
(2)
In‐vessel thermal and hydraulic analysis code
Coolant flow rate and coolant temperature distribution
(3)
TEMDIM
Fuel temperature analysis code Fuel temperature
5
In‐vessel Thermal and Hydraulic Analysis Code “FLOWNET”
: Gap between each block
: Control rod column flow path : Fuel channel
Top shielding
Replaceable reflector
Fuel assembly
Replaceabl
e reflector
Hot plenum




One‐dimensional model using nodes and branches The flow channels are represented by node, and the nodes are connected by branch.
The heat transfer between the branches are taking into account.
Flow paths: the main coolant flow, the bypass flow in the inter‐column gaps, the leakage flow through the permanent reflectors and the cross flow in the horizontal interface gaps of the hexagonal graphite blocks
Ref: S. Saito et al., “Design of High Temperature Engineering Test Reactor (HTTR)”, JAERI 1332 (1994).
6
Fuel Temperature Analysis Code “TEMDIM”
 Two dimensional cylindrical model
 based on the power distribution including local power peaking, coolant flow distribution including redistribution in the fuel column and hot spot factors
Fundamental equation
T
Gas
in
Coolant
5
  Fi  ∆Ti 
i 1
n(i)
m(i)


2 
Fi   fs i, j 1   fri,k 
k 1


j 1
T
FUEL
Tgin
T
N
N
N
N
T2 T3 T4 T5
N
T1
N
T FUEL: Fuel temperature (℃)
A‐A
cross
section
TinGas : Coolant inlet temperature (℃)
∆T i : Temperature rising (℃)
Fi
: Hot spot factor (‐)
Estimated
A point
A
fs i, j : Random factor (e.g., manufacturing tolerances, flow rate, inlet coolant temperature )(‐) i= 1 : Coolant temperature rising
2 : Film temperature rising
3 : Temperature rising in graphite sleeve
4 : Fuel compact‐graphite sleeve gap temperature rising
5 : Temperature rising in fuel compact
Fuel rod
fri,k : Systematic factor (e.g. total reactor power, coolant Graphite block
uncertainties on physical properties) (‐) Fuel compact
Gap
Graphite sleeve
Annular flow path
Coolant
7
Evaluation Result of Fuel Temperature at Design Stage Top plenum
Upper
shield
Fuel
block
Replaceable
reflector
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐
Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐
67(2006).
5.5
92.2
1.3
5.5
92.1
1.4
5.7
7.2
91.6
88.4
1.8
0.4
3.7
88.8
1.7
8.2
90.1
1.2
7.1
91.4
1.0
8.8
99.3
Plenum
block
Inner column gap
Replaceable
reflector
Fuel cooling channel
 Remains are the leakage flow through the permanent reflectors, flow in the control rod cooling channel, and the bypass flow in the inner column gap. 5.5
Control rod cooling channel
 The effective flow rate for the fuel cooling is about 88% of total flow. 98.9
0.4
0.7
Hot plenum
(Unit : %)
100 Outlet Inlet 100
8
Evaluation Result of Fuel Temperature at Design Stage Horizontal temp. distribution
Nominal temp.
Maximum temp.
Vertical temp. distribution for the fuel where the maximum fuel temp. appeared.
Temperature (oC)
1293
1321 1492
1475 1302 1492
1476
Core Center
Vertical position at a fuel column
1305
1293 1474 1295
1476
1295 1475
1305 1476 1302
1474
1476 1321
400
(Top)
600
800 1000 1200 1400 1600
1
2
3
4
Graphite block
Coolant
5
Sleeve outer surface
(Bottom)
Compact inner surface (nominal) (maximum)
Temperature limits of HTTR fuel
 Anticipated operation occurrence: 1600 ˚C
 Normal Operation : 1495˚C
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
9
Re‐evaluation of HTTR Fuel Temperature
(1) 1st modification using the operation data of the rise‐to‐
power test up to 850˚C
 Validation of power and helium flow distributions
 Revision of operating conditions (e.g., core inlet coolant temperature)
 Revision of hot spot factors
(2) 2nd modification using new analysis model
 Detailed mesh model
(3) 3rd modification using the 950˚C operation data in 2004
 Revision of core inlet coolant temperature and core coolant flow rate
10
Validation of Evaluation Results using HTTR data Estimated power and helium flow distributions were validated using the operation data of the rise‐to‐power test up to 850oC. Power distribution
Flow distribution
F5
F4
T/C for core inlet coolant temp.
F3
F3
F5
F2
F4
F1
F1
F2
Center region
Center
Outside region
(6 points)
The gross gamma ray from the fuel assemblies was measured by GM counter
1.00
0.75
Measured
Calculated
0.50
F1
F2
F3,F4
Column number
F5
Coolant temperature rising (oC)
Power density (W/cc)
1.25
T/C for hot plenum coolant temp.
490
480
470
460
450
440
Measured
Calculated
Center region
Outside region
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
11
1st Modification: Modification of Calculation Conditions
Boundary conditions concerning to the operating conditions, and hot spot factors (i.e., systematic factors) were revised using the operation data up to 850oC operation. Design
Revised
Reasons
Operation day
440 EFPD
160 EFPD
Operating condition at 1st 950oC operation
Core inlet coolant temperature
415oC
409oC
Core coolant flow rate
10.2 kg/s
10.1 kg/s
Re‐evaluation of heat and mass balance using operation data
Control rod position
2610 mm
2900 mm
Operation data
Coolant temp. rise
2.5%
0%
Others
2.5%
2.0%
4.0%
0%
4.0%
2.0%
3.2%
1.6%
Factor for thermal
power
Systematic factors
Factor for axial power distribution
Factor for Coolant temp. rise
flow distribution Film temp. rise
Operation data
No effect of the thermal power error on the coolant temp. rise
Calibration result of thermal power
Measurement results of power distribution
Re‐evaluation of flow distribution using operation data Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
12
1st Modification: Re‐evaluation Result of Fuel Temperature Top plenum
Upper
shield
Replaceable
reflector
92.2
1.3
5.5
92.1
1.4
5.7
7.2
8.8
91.6
88.4
Replaceable
reflector
0.4
3.7
1.7
8.2
90.1
1.2
7.1
91.4
1.0
0.4
Fuel
block
Replaceable
reflector
0.7
Hot plenum
(Unit : %)
6.2
1.8
88.8
99.3
Plenum
block
Inner column gap
Fuel
block
5.5
Fuel cooling channel
Replaceable
reflector
Control rod cooling channel
5.5
100 Outlet Inlet 100
99.6
6.2
90.5
1.7
6.2
90.4
1.7
6.2
6.3
6.0
90.2
89.6
1.9
0.2
2.4
90.3
2.1
5.2
91.4
2.1
5.0
92.9
0.9
99.2
Plenum
block
Inner column gap
Upper
shield
98.9
Control rod cooling channel
Top
plenum
1st modification
Flow distribution
Fuel cooling channel
Design stage
0.5
0.2
Hot plenum
(Unit : %)
100 Outlet Inlet 100
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
13
1st Modification: Re‐evaluation Result of Fuel Temperature Fuel temperature
Design stage
Nominal temp.
Maximum temp.
1st modification
Nominal temp.
Maximum temp.
1305
1293 1474 1295
1476
1295 1475
1305 1476 1302
1474
1476 1321
1307
1310 1463 1311
1452
1311 1452
1307 1453 1291
1463
1434 1286
1293
1321 1492
1475 1302 1492
1476
1310
1286 1431
1452 1302
1291 1431
1476
1434
Core Center
Core Center
Estimated maximum fuel temperature was decreased from 1492oC at design stage to 1463oC by the re‐evaluation using the operation data. Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
14
1st Modification: Re‐evaluation Result of Fuel Temperature Temperature distribution
1st modification
Design stage
600
1400
Vertical position at a fuel column
2
3
Graphite block
Coolant
5
600
Temperature (oC)
800 1000 1200
1400
1600
(Top)
1
(Top)
1
4
400
1600
Sleeve outer surface
(Bottom) Compact inner surface (nominal)
(maximum)
Vertical position at a fuel column
400
Temperature (oC)
800 1000 1200
2
3
Graphite block
4
Coolant
Sleeve outer surface
5
(Bottom) Compact inner surface (nominal)
(maximum)
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
15
2nd Modification: Modification of Analysis Model
Old model
 Each fuel block is divided into 6 triangular‐
meshes for the models of nuclear design and
the fuel temperature analysis.
 One fuel rod is represented by the triangular‐
mesh of the fuel temperature analysis model .
 The rector power calculated by the nuclear
design is allocated for each mesh of the fuel
temperature analysis model using the peaking
factor.
Allocated rector power
×hot spot factor
 The reactor power is multiplied by the hot
spot factor to take into account the
heterogeneous effect of the nuclear design
model (i.e., power: +7%).
New model
The hot spot factor related to the heterogeneous effect of the nuclear design model is eliminated by using detailed mesh.
16
2nd Modification: Modification of Analysis Model
Old model
New model
Horizontal
1/6‐divided‐block model 1
Vertical
2
Fuel rod
3
4
4 meshes
Each fuel rod model
1
2
3
4
5
6
7
8
9
10
11
12
13
14
14 meshes
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
17
2nd Modification: Modified Evaluation Method
(1)
Nuclear design code
Added
Power density and neutron fluence distributions
(3)
Fuel, control rod, core component, core internal structure design data
FLOWNET
(2)
In‐vessel thermal and hydraulic analysis code
Coolant flow rate distribution
MVP
Continuous energy Monte Carlo code Power distribution for each fuel rod
(4)
TEMDIM
Fuel temperature analysis code Fuel temperature
18
2nd Modification: Re‐evaluation Result of Fuel Temperature Fuel temperature
Old model (1st modification)
New model (2nd modification)
Nominal temp.
Maximum temp.
1307
1463
1310
1452
1307
1310 1463 1311
1452
1311 1452
1453 1291
1434 1286
1286 1431
1302 1431
1291
1476
1434
1256
1368
1261
1382
1258
1251 1371 1255
1377
1265 1372
1387 1314
1428 1214
1221 1322
1302 1330
1303
1476
1417
Old model
New model
Temp. difference
Maximum fuel temp.
1463oC
1428oC
‐35oC
Core average fuel temp.
1178oC
1018oC
‐160oC
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
19
2nd Modification: Re‐evaluation Result of Fuel Temperature Temperature distribution
Old model (1st modification)
600
1400
Vertical position at a fuel column
600
Temperature (oC)
800 1000 1200
1400
1600
(Top)
1
(Top)
1
2
3
Graphite block
4
400
1600
Coolant
Sleeve outer surface
5
(Bottom) Compact inner surface (nominal)
(maximum)
Vertical position at a fuel column
400
Temperature (oC)
800 1000 1200
New model (2nd modification)
2
3
Graphite block
4
Coolant
Sleeve outer surface
5
(Bottom) Compact inner surface (nominal)
(maximum)
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
20
3rd Modification: Re‐evaluation Result of Fuel Temperature Analysis result
(2nd modification)
Operation data at 950oC operation in 2004
Coolant temp. (oC)
Core inlet
Center region
399
396
Outside region
405
402
Center region
984
991
Outside region
952
954
Center region
585
595
Outside region
547
551
1,428
Analysis result using operation data (3rd modification)
Core outlet
Temperature rising
Maximum fuel temp. (oC)
1,435
Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).
21
Fuel Temperature Measurement (Future Test)
Objectives
 To develop the fuel temperature measuring technique for the prismatic HTGR
 To upgrade the core design technology for the prismatic HTGR
Methods
 Temperature monitors using 22 kinds of melting wire in temperature range 600 ‐1390oC
 Irradiation performance of the melt‐wire temperature monitor has been confirmed by JMTR capsule irradiation test.
 Temperature monitors will be inserted into the fuel assembly to measure temperature distribution of the HTTR core.
HTTR
Melt‐wire temperature monitor
Monitors will be stacked into hole under the dowel pin of the fuel assembly
Fuel assembly
Ref: Y. Tachibana et al., “Test Plan using the HTTR for Commercialization of GTHTR300C”, JAEA‐Technology 2009‐063 (2009).
22
High Temperature Irradiation of the HTGR Fuel
/ Metallic FP Plate‐out Test (Future Test)
Objectives
 To optimize the limitation of fuel failure under accident condition
 To investigate metallic FP (Sr, Cs, etc.) plate‐
out behavior by using the real HTGR facility
Melt‐wire temp. monitor
Methods
 A test fuel element loaded at the center column of HTTR and heated step‐by‐step up to 2000oC
 Using temperature monitor to measure fuel temperature
 On‐line measurement of primary coolant radioactivity to estimate additional fuel failure fraction
 Plate‐out probe( ) will be settled in the primary circuit to measure metallic FP plate‐
out concentration by PIE
 PIEs to investigate fuel failure and FP plate‐
out Ref: Y. Tachibana et al., “Test Plan using the HTTR for Commercialization of GTHTR300C”, JAEA‐Technology 2009‐063 (2009).
Test fuel element
Reactor
to SPWC
IHX
By-pass line
For
parallelloaded
operation
HGC
to
auxiliary
cooling
system
For
singleloaded
operation
HGC
PPWC
To pressurized water cooling system
23
Summary
 HTTR fuel temperature was re‐evaluated using the HTTR operating data and new analysis model.  The summary of the re‐evaluation results of the maximum fuel temperature is the following:




Design stage
1st modification using 850˚C operation data
2nd modification new analysis model
3rd modification using 950˚C operation data
: 1492oC
: 1463oC : 1428oC
: 1435oC
 We are planning to measure the HTTR fuel temperature using the melt‐wire temperature monitor. 24