29.05.2014 1 Thor Energy Th/Pu test (IFA-730): In-pile data from the first irradiation cycle Presented by C. Vitanza, IFE/Halden
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29.05.20141
Thor Energy Th/Pu test (IFA-730): In-pile data from the first irradiation cycle
Presented by C. Vitanza, IFE/Halden
The Halden Reactor - HBWRT2
T1
S61
10
S53
S30 S35 S47 S15
504 S39 S26 16 E06 17 DO4 S65
S60 31 S33 647 32 654 S49 33 S28
515 648 22 E01 E02 23 D01D02
552 S23 21 S45 S38 11 638 603 D10 S36616
642 30 641 S75 S57 D11 12 604 S17 34 S50
35 S52 640 20 602ECCS 557 25 S32 597 24
S63 39 626 548 15 590 628 S34 E04 645 D09
S54 S51 29 S22 630 14 503 E08 26 E05 S81
S48 S42 S44 28 617 639 27 644 E07 S62 S58
S82 S64 38 S40 636 37 637 D03 36 S21
633 S55 19 S76 S18 18 595 625
13 651 652 S46 S78 D08 S79
S73 S83
S31
D05 610
S56
S77
S27 S84
S80 S59
D06
E03 C04
S37
D07 C03
• 20 MW heavy water reactor at 34 bar and 235oC
• Height of active core is 80 cm• Hexagonal array of 300
channels within pressure tank• Rings 1-6 used for
experiments (about 30/110 positions used)
• Experimental channels directly in HBWR (71 mm in diameter)
• 8-10 loops system in operation at any given time (32 to 45 mm in diameter)
Instrumented Test RigsIn-core Connector
Outlet Coolant Thermocouples
Fuel Centre - line Thermocouple
Neutron absorber
Pressure transducer (PF)
He-3 coil
Neutron Detector (V-type)
Differential Transformer (LVDT)
Inlet Coolant Thermocouples
Inlet Turbine Flowmeter
Shroud (Ø 73/71mm)
Calibration Valve
Fuel Rod storageposition
Fuel Rod rampposition
� Several fuel rods can be tested at the same time, hence enabling to compare different fuel rod variants, such as cladding type, fuel pellet type, pellet-cladding gap etc.� On-line instrumentation provides
relevant information on the fuel rod performance, e.g. on PCMI and rod integrity (and also FGR if needed )� Additional information on other
performance aspects can be obtained from PIE (dimensional & corrosion)� The irradiation is carried out at constant
power and can last for a long time (one or more years). After that, one or some of these rods can be subjected to special testing, e.g.
power ramp or even LOCA.
Rod behavior in normal operation
Integrity of fuel rods at base irradiation conditions
4
Temperature histories• Peak fuel temperatures were close to fission gas release threshold temperatures during the 1st and 2nd interlinkage tests.
The 1st The 2nd
Rod pressure (fission gas release)
FGR (reactor power ramp to 20 MW)
Signs of densification in rods 1 and 2Different behaviour for 8 %wt Gd because T < Tthreshold
Integral test: Fuel pellet densification and swelling
Rod 1
Rod 2
Rod behavior in normal operation
Cla
ddin
g s
trai
n (
PC
MI)
Example of failed rod
Large PCMIbut no failure
Integrity of fuel rods in power ramp conditions Assessment of PCI and PCMI margin
8
• Single rod experiment using high burnup fuel• Heating provided from within the rod by low
level of nuclear power simulating decay heat
• 3 cladding TCs• 3 heater TCs• Fuel pressure transducer• Cladding elongation detector
Our previous experience: IFA-650
Heater cable
Ø 34 Flask
Ø 9.5 rod
Ø 26.5 /Ø 20 heater
HeaterT/C
9
29.05.2014
IFA-730.1 Status• Rig operated since 28.4.2013
• ~150 operation days so far• Power kept at ~30 kW/m for large diam. rods
~20 kW/m for small diam. rods• Measured temperature kept below ~1200°C
• In order to keep long-term operation of the TFs• Power calibration on 28.4.
• Good agreement with pre-calculation (±3%)
• Behaviour of Th fuel rods as expected
29.05.201411
IFA-730Test Matrix – Phase 1a
5/29/2014
U
TF
U/Th
EC
Pu/Th
U/Th
TF
U/7%Th
PF
TF
U/Th
PF
Pu/Th
U/Th
TF
U/7%Th
EC
TF
PF
TF
U/Th
PF
PF – Pressure transducer
TF – Fuel thermocouple
EC – Cladding elongation detector
5/29/2014
IFA 730 test rig
IFA-730.1 Instrument status• From 29.4. Rod pressure behaviour in UO2 rod not as
expected• Early on only weak or no response to power changes
• Mechanical friction • Strong drop on 3.5.• After this, the instrument shows reasonable response to power and
temperature changes and is likely to show clearly when FGR occurs
• On 3.5. a jump of ~50°C in TF2• Signal check showed no indication of failure• Subsequently following operation as expected
• Sudden drop in PF5 signal on 3.6. • Signal check showed water penetration into the cable• Instrument faulty
29.05.201414
29.05.201415
IFA-730 Test matrixRod ID 730-1 730-2 730-3 730-4 730-5 730-6
Fuel 58% U / 42% Th8% Pu / 92%Th(OMICO pellets)
93% U / 7% Th 58% U / 42% Th8% Pu / 92%Th(OMICO pellets)
93% U / 7% Th 58% U / 42% Th UO2
Pellet OD [mm] 5.90 8.48 5.90 8.48 5.90 8.48
Diam. gap [ µm] 125 150 125 150 125 150
Instr. TF / EC TF1) / PF TF / PF TF / EC TF / PF2) TF / PF3)
Power [kW/m] 20 32 20 30 20 32
Burnup [MWd/Ox] 6.2 3.9 6.0 3.8 6.1 3.9
1) Thermocouple TF2 showed a jump on 3.5. – continued working normally.2) Pressure transducer PF5 faulty since 3.6. – wet cable3) Pressure transducer PF6 unreliable at the beginning – likely to show FGR
IFA-730 Power levels
16
IFA-730 Fuel temperature
29.05.201417
• Expectation:• Temperature in ThU rods lower or similar to reference UO2 � confirmed
• Temperature in PuTh pellets lower than in reference ThU � confirmed
• Further irradiation will provide information on long term behaviour
IFA-730 Results: Fuel temperature• Comparison with model predictions, start-up
data: large diameter rods• In general good agreement between measured temperature and
model predictions
• Temperature in Th fuel slightly lower than reference UO2 fuel
29.05.201418
IFA-730 Results: Fuel temperature• Comparison with model predictions, start-up
data: small diameter rods• In general good agreement between measured temperature and
model predictions
• Temperature in Th/Pu fuel slightly lower than Th/U fuel
29.05.201419
IFA-730 Normalised temperature 1/2
20
IFA-730 Fuel rod pressure
29.05.201421
IFA-730 Cladding elongation
22
IFA-730.1 Summary 1/2• Rig operated since 28.4.2013
• Power kept at ~30 kW/m for large diam. rods~20 kW/m for small diam. rods
• Collection of irradiation data on • Fuel centre temperature Fuel thermal conductivity• Rod pressureFuel dimensional stability and fission gas release• Cladding elongation Pellet-cladding mechanical interaction (fuel dimensional stability)
• Continued operation planned at current power levels
29.05.201423
IFA-730.1 Summary 2/2• Data collected so far indicates that Th fuel behaves
according to expectation• Fuel centre temperatures during first start-up are in fair
agreement with model predictions for fresh fuel • Temperatures in Th fuel lower or similar to reference UO2
• As expected• Further irradiation will provide information on long term behaviour
• Data collected during the further irradiation may provide information about fission gas release behaviour (rod pressure)
• Current data will be analysed more carefully during the upcoming reactor outage
29.05.201424
Thermal conductivity of Pu/Th
29.05.201425
• C. Cozzo et al. , Journal of Nuclear Materials 416, 135-141, 2011
Thermal conductivity of U/Th
29.05.201426
• K. Bakker, Journal of nuclear materials 250, 1-12, 1997.• J.H. Yang, Nuclear Technology 147, 113-119, 2004.
Thermal conductivity of U/Th
29.05.201427
• K. Bakker, Journal of nuclear materials 250, 1-12, 1997.• J.H. Yang, Nuclear Technology 147, 113-119, 2004.
IFA-730Test Matrix – Phase 1a
5/29/2014
U
TF
U/Th
EC
Pu/Th
U/Th
TF
U/7%Th
PF
TF
U/Th
PF
Pu/Th
U/Th
TF
U/7%Th
EC
TF
PF
TF
U/Th
PF
PF – Pressure transducer
TF – Fuel thermocouple
EC – Cladding elongation detector
IFA-730Test Matrix – Phase 1b
5/29/2014
PF – Pressure transducer
TF – Fuel thermocouple
EC – Cladding elongation detector
U
TF
U/Th
EC
Pu/Th
U/Th
TF
Pu/Th
PF
TF
U/Th
PF
Pu/Th
U/Th
TF
EC
TF
PF
TF
U/Th
PF
Pu/Th
5/29/2014
IFA 730 program, an opportunity for Japan
� The test is performing well� Unique data are being produced and � Will continue to be produced on Th/Pu fuel� An international Consortium led by TE is in place
Joining this Consortium will be the most efficientand cost effective way for Japan-nuclear to get access to unique data and to be part of the Thorium future
29.05.201431
� Single rod experiment using high burnup fuel� Heating provided from within the rod by low
level of nuclear power simulating decay heat
� 3 cladding TCs� 3 heater TCs� Fuel pressure transducer� Cladding elongation detector
In-reactor simulation of LOCAFeatures of the Halden LOCA rig
Heater cable
Ø 34 Flask
Fuel rod
Ø 26.5 /Ø 20 heater
HeaterT/C
32
Rod behavior in transient operation
Appearance of rod cladding after LOCA (Halden test 9)(rod probably broken during test)
� Large balloon at bottom of rod
� Complete circumferential crack
� Hydrogen content near crack ~1600 ppm
Rod behavior in transient operation
33
Example 6: Tolerable rod overpressure in UO2 and MOX fuel rods
Fuel rod
Booster rods
Outletthermocouple
Inletthermocouple
Pressure flask
Gas line
Gas line
Fuelthermocouple
• PWR operation conditions• Pressure flask surrounded by
12 booster rods• Gas flow lines
• Overpressure• Hydraulic diameter measurement
and gamma-spectroscopy
• Thermocouple (TF) with in-core connector
• Cladding extensometry (EC)
Tolerable Rod Overpressure Tests
Overpressure, bar
Tem
pera
ture
incr
ease
rat
e at
15
kW/m
°C
/ 10
00 h
ours
100 (+155) bar
Observations• Rate of temperature
increase correlated with overpressure
• Thermal feedback occurs only at considerable overpressure (>100 bar)
• Below this threshold, clad creep-out is sufficiently compensated by fuel swelling, and no net thermal feedback becomes apparent
IFA-730 Normalised temperature 2/2
29.05.201436
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