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Behavior of high burnup fuel during LOCA - Key observations and test plan at JAEA -
Fumihisa Nagase
Japan Atomic Energy Agency
IAEA Technical Meeting on Fuel Behaviour and Modellingunder Severe Transient and LOCA Conditions
Ibaraki, JapanOct. 18-21, 2011
1
Contents
• Whole view of LOCA study at JAEA• Main results on oxidation behavior• Main results on thermal shock resistance• New program with advanced cladding
2
Whole view of LOCA study at JAEA
• Sample– Hydrided, unirradiated cladding (Zircaloy)– 39 − 44 GWd/t fuel cladding (Zircaloy)– 66 − 76 GWd/t cladding (Improved alloys)
• Test method– Oxidation rate measurement– Semi-integral quench test– Mechanical test (Ring-compression, ring-tensile,
axial tensile and bending)– Separate test on secondary hydriding, restraint load,
etc.
3
Oxidation kinetics evaluation in a wide range
100
101
102
103
104
105
106
10710
-2
10-1
100
101
102
1573K1523K
1473K1423K1373K
1323K1273K
1253K
1223K
1173K1123K
1073K
973K
873K
773K
Wei
gh
t g
ain
( m
g/c
m2 )
Oxidation time ( s )
Results for low-tin Zircaloy-4- Parabolic rate law< 3,600 s from 1,273 to 1,573 K< 900 s from 773 to 1,253 K- Cubic rate law
> 900 s from 773 to 1,253 K
- The Baker-Just equation not conservative below 1,073 K.
- Breakaway oxidation for longer periods at specific temperatures .
Breakaway
Effect of pre -oxidation – Unirradiated cladding -
metal
120 s 180 s 900 s 1800 s
Uniform growth
Oxidation initiated at cracking position
Pre-formedoxide
HToxide
Weight gain is smaller in the pre-oxidized cladding than in the as-received one.
Protective effect of the pre-formed oxide for high temperature oxidation
Low-tin Zry-4
100µm
4
Oxidation time (s)
Wei
ght g
ain
(mg/
cm2 )
Influence of pre-hydriding varies depending on oxidation temperature and hydrogen concentration.
The influence is estimated to be small under the postulated LOCA conditions (<800 ppm, <900 s).
Unirradiated low-tin Zry-4
Effect of pre -hydriding - Unirradiated cladding -
5
Oxidation test with high burnup fuel cladding
M5 ZIRLO Zry-2
Test specimens were prepared from high burnup fuel rods, irradiated at European power plants.
Cladding ReactorLocal Burnup
(GWd/t)Corrosion layerthickness (µm)
hydrogencontent (ppm)
M5 Ringhals 68 6-7 60-80
ZIRLO Vandellos 79 84-98 900-990
Zry-2(LK3) Leibstadt 73 20-33 170-180
6
Test Apparatus and Test Conditions
Steam supply rate: 30 mg/s Test method: two-sided oxidationTemperature: an R-type Thermocouple at the specimen holder
M5, Zry-2 1000, 1100, 1200oCZIRLO 900, 1050, 1200oC
Time: 120 to 4000sMeasurement: weight gain and oxide layer thickness
T.C
Electrical resistance furnace
T.C
Specimen
Steam GeneratorTemperatureControl
Recorder
Reaction tubeSpecimen
7
- The weight gain of the irradiated specimens is smaller than that of the unirradiated samples.
- The difference becomes smaller at higher temperatures.
- The oxidation of the irradiated ZIRLO cladding approximately obeys a parabolic rate law.
∆W 2 = Kw .t ∆W : weight gain
Kw : parabolic rate constantt : oxidation time
Weight Gain of irradiated ZIRLO Cladding
8
Metallography of irradiated ZIRLO Samples after oxi dation test
- The HT oxidation initiated at cracking positions of the corrosion layer on the cladding outer surface. The oxide layer uniformly grows at higher temperature or after longer oxidation times.
- Oxide layer is uniform on the cladding inner surface.
1173 K, 4022s
HT oxideHT oxide
1323 K, 1801s 1473 K, 601s
HT oxide
Corrosionlayer
OD
50µm
Magnified
HT oxide
9
- The outer surface oxide is thinner in the irradiated Zry-2 specimens at temperatures <1200oC.
- Growth of the inner surface oxide in the irradiated cladding is almost equivalent to that in the non-irradiated Zry-4.
Oxide Layer Growth in Irradiated ZIRLO Cladding
Protective effect of the corrosion layer
Oxi
de la
yer
thic
knes
s (m
m)
10
11
BJ eq.
CP eq.
Parabolic Rate Constants for Weight Gain
6.5 7 7.5 8 8.5Reciprocal temperature (x10-4/K)
Par
abol
ic r
ate
cons
tant
, Kw
(g2 /
m4 /
s)
100
101
102
10-1
Temperature (oC)
ZIRLO
non-irradiated M5MDAZry-4 - Oxidation rates of the examined
alloy cladding are lower than that given by the Baker-Just equation.
Zry-2
irradiated
900100011001200- Rate constants of the irradiated
cladding are lower than those of the non-irradiated Zry-4 at 1100oC and below, while difference is small at 1200oC.
11
12
JapanBased on fracture/no-fracture boundary determined by the semi-integral quench test that simulates the LOCA sequences such as ballooning, rupture, oxidation, thermal shock by reflooding and mechanical loading.
U.S. and European countriesBased on Zero-ductility criteria determined by the ring-compression test of oxidized cladding.
Base of ECCS acceptance criteria
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Semi-integral quench test
• Test rod experiences the whole process during a LOCA, ballooning and rupture, oxidation and hydrogen absorption at high temperature, and thermal shock at quenching.
• The rod is quenched with axial constraint to represent a possible condition of fuel rods between grid positions.
Quartz tube
Grabbing device
Steam outlet
Steam inlet Flooding water inlet
Infrared furnace
Test rod
Restraint load control system
Equipped on tensile testing machine
Load cell
Cla
ddin
g le
ngth
:19
0 to
600
mm
TC
Wel
ded
Wel
ded
Cla
ddin
g le
ngth
:19
0 to
600
mm
TC
Wel
ded
Wel
ded
0
400
800
1200
1600
2000
0
200
400
600
800
1000
0 400 800 1200
No.4No.4
No.2
No.3
No.1 TC
Tem
pera
ture
(K
)
Axi
al L
oad
(N)
Time (s)
Rupture
Temperature
Load
14
Main results from experimentswith unirradiated cladding
• Decreases of rupture temperature and circumferential burst strain were observed with an increase in the initial hydrogen concentration, which might be associated with the reduction of cladding strength and the shift of phase transformation temperatures.
• The cladding fracture during quench was primarily dependent on the amount of oxidation. The fracture threshold for the amount of oxidation was reduced by an increase in the initial hydrogen concentration and the axial restraint load.
• An increase in hydrogen concentration increases the precipitation of fine hydrides and the oxygen concentration in the prior phase region, which causes of enhanced embrittlement of oxidized claddings with a higher initial hydrogen concentration.
15
High burnup fuel cladding samplesand test conditions
• Original fuels: PWR (Zirlo, M5, MDA, NDA), BWR (Zry-2)< 76 GWd/t (local)
• Oxidation temperature: 1463−1480 K (1190−1207°C)• Oxidation time: 131−719 s• Oxidation amount: 18.3−38.0% (16.4−35.6%) ECR
Calculated with the Baker-Just equation with oxidation temperature and time, for reduced metallic thickness after ballooning. The value in parenthesis is ECR for the initial metallic thickness before ballooning.
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SegmentID
Burnup(GWd/t)
CladdingCorrosion
layer(mm)
Initialhydrogen(wppm)
MDA1R76 MDA
48 720
MDA2R 60 838
MFI166 M5
6 73
MFI2 7 69
NDA1 69 NDA 32 214
ZIR2R 71ZIRLO
50 496
ZIR3R 73 79 764
ZRT1 66Zry-2
30 297
ZRT2 73 24 182
High burnup fuel cladding for se-mi-integral tests
MDA: Zr-0.8Sn-0.2Fe-0.1Cr-0.5NbNDA: Zr-1.0Sn-0.27Fe-0.16Cr-0.1Nb-0.01Ni
17
Post-test appearances and cross section
Radial cross sectionat rupture position Rupture
MDA-1R Rupture openingRupture opening
MFI-2 Rupture openingRupture opening
ZIR-2R Rupture openingRupture opening
NDA-1 Rupture openingRupture opening
Fracture positionFracture positionMDA-2R
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SegmentID
BurstTemp.(°C)
BurstStrain(%)
OxidationTemperature
(°C)
OxidationTime(s)
Oxidationamount(%ERC)
Fracture/no-
fracture
Restraintload(N)
MDA1R 715 9.9 1207 131 18.3 N.F. <350
MDA2R - 6.0 1190 719 38.0 F. 530
MFI1 780 20.1 1197 151 19.5 N.F. <400
MFI2 762 19.2 1196 229 23.6 N.F. ~0
NDA1 715 8.9 1194 280 22.5 N.F. 518
ZIR2R 672 28.0 1200 228 27.3 N.F. 518
ZIR3R 676 20.6 1186 153 20.2 N.F. 519
ZRT1 763 17.7 1195 222 21.2 N.F. 519
ZRT2 778 20.6 1194 232 22.0 N.F. 0
ECR is calculated with the Baker-Just equation with oxidation temperature and time, for reduced metallic thickness after ballooning. The value in parenthesis is ECR for the initial metallic thickness before ballooning.
Summary of test results
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Fracture map
ECR was calculated with the Baker-Just equation with oxidation temperature and time, for reduced metallic thickness after ballooning.
1400 1450 15000
10
20
30
40
50
Oxidation temperature (K)
Cal
cula
ted
EC
R (
%)
Japanese safety criteria:<1473 K<15% ECR
Fracture during quenchNo-Fracture
18.3 to 27.3%
38.0%
0
10
20
30
40
50
0 200 400 600 800 1000
Oxi
datio
n (%
EC
R)
Initial hydrogen concentration (ppm)
MFI-1(<400N)
MFI-2(0N)
MDA-1R(<350N)
ZIR-3R
MDA-2R
ZIR-2R
ZRT-1ZRT-2(0N)
Fracture boundary of unirradiated Zry-4
Safety limit(15%ECR)
Fracture No-fracture66-77GWd/t
0
10
20
30
40
50
0 200 400 600 800 1000
Oxi
datio
n (%
EC
R)
Initial hydrogen concentration (ppm)
MFI-1(<400N)
MFI-2(0N)
MDA-1R(<350N)
ZIR-3R
MDA-2R
ZIR-2R
ZRT-1ZRT-2(0N)
Fracture boundary of unirradiated Zry-4
Safety limit(15%ECR)
Fracture No-fracture66-77GWd/t
The fracture boundary of the high burnup cladding is higher than the 15% limit in the Japanese ECCS acceptance criteria.
The fracture boundary is not reduced significantly by high burnup and use of new alloys in the examined burnup level, though it may be somewhat reduced with pre-hydriding as observed in unirradiated Zry-4 cladding.
20
Ring compression testsafter semi-integral quench test
Objective
To obtaining a wide range of database for the safety of high burnup fuel and compare the two test methodologies, which are used for the regulatory judgments
Example of sampling
0
50
100
150
200
-0.5 0 0.5 1
Load
(N
)
Displacement (mm)
LMDA1R
RC1
RC2
Load -displacement curves and post-test appearances of MDA cladding
Cladding Specimen Hydrogen Ox. Temp. ECR (B-J) ECR (C-P)
MDA-1RRC1 870 ppm 1158ºC 13.4% 10.6%
RC2 750 pp 1137ºC 11.8% 9.5% 21
Plastic strain at failure as function of oxidation amount and hydrogen concentration
• The plastic strain to failure decreases with the oxidation and the hydrogen concentration.
• Better correlation is seen in the hydrogen dependence. Ductility obviously decreases when the hydrogen concentration is above 500ppm.
• Most the examined specimens fractured without plastic strain, though the high burnup cladding did not fracture on quenching in the integral thermal shock tests. 22
23
Subjects to be studied
• Behavior and fracture under LOCA conditions of high burnup cladding with advanced alloys
• M-MDA, Opt. Zirlo and Zr-Nb binary alloys including M5• Loading conditions including restraint during quench• Secondary hydriding
• High burnup effect• Influence of temperature transient
• Tests for long-term cooling• e.g. Mechanical tests considering possible loading during
in-core and out-core storage, handling, etc.• Tests for beyond DBA-LOCA in consideration with the
accidents at Fukushima-Daiichi
24
Claddingmaterial
Burnup Reactor
PWR fuel cladding
M-MDA 73GWd/tVandellos-2
(Spain)
M5 78GWd/tGravelines-5
(France)
Low-tin ZIRLO 73GWd/tVandellos-2
(Spain)
BWR fuel cladding
Zry-2/LK3 92 GWd/tLeibstadt
(Switzerland)
Sample and test conditions in new program
25
Test conditions
• Oxidation test• Oxidation temperature: 1223−1473 K (950−1200°C)• Oxidation time: 120−3600 s• Post-test analysis: Weight gain measurement,
Microstructure observation and Oxide layer thickness measurement
• Integral thermal shock test• Oxidation temperature: 1423−1473 K (1150−1200°C)• Oxidation amount: 15−30% B-J ECR• Restraint condition: to be determined
(Basically 540 N, but may be reduced depending on results of the evaluation in future)
26
LOCA test schedule
Jp FY(Apr-Mar)
2009 2010 2011 2012 2013 2014 2015
Shipment
Oxidation test
Integral thermal shock test
Inside EuropeTo Japan
Unirradiated cladding
High burnup cladding
Unirradiated cladding
High burnup cladding
27
Conclusion
• JAEA has been conducting LOCA studies with unirradiated and irradiated fuel cladding. As a result, various information has been obtained on cladding oxidation, ballooning and rupture behavior, cladding embrittlement and condition of fracture on quench of high burnup fuel under LOCA conditions.
• JAEA is proceeding with the second phase of the experimental program with high burnup fuel cladding irradiated European power reactors. It is expected that the experiments provide information for future regulation of high burnup fuels which use advanced cladding alloys.
• More investigations on secondary hydriding, restraint loading and long term cooling are required for better understanding of the high burnup fuel behavior under LOCA conditions.
• Tests for beyond DBA-LOCA would be necessary in consideration with the accidents at Fukushima-Daiichi power plant.
28
The experimental program with the European high burnup fuel is conducted as part of a program sponsored and organized by Nuclear and Industrial Safety Agency (NISA), Ministry of Economy, Trade and Industry (METI).
Acknowledgment
29
References• F. NAGASE, T. OTOMO and H. UETSUKA, “Oxidation Kinetics of Low-Sn Zircaloy-4 at the
Temperature Range from 773 to 1573 K,” J. Nucl. Sci. Technol., 40[4] (2003) 213-219.
• T. CHUTO, F. NAGASE and T. FUKETA, High Temperature Oxidation of Nb-Containing Zr Alloy Cladding in LOCA Conditions, Nuclear Engineering and Technol., 41[2], 2009, p.163-170.
• Y. UDAGAWA, F. NAGASE and T. FUKETA, “Effect of Cooling History on Cladding Ductility under LOCA Conditions,” J. Nucl. Sci. Technol., 43[8] (2006) 844-850.
• F. NAGASE and T. FUKETA, “Effect of Pre-Hydriding on Thermal Shock Resistance of Zircaloy-4 Cladding under Simulated Loss-of-Coolant Accident Conditions,” J. Nucl. Sci. Technol., 41[7] (2004) 723-730.
• F. NAGASE and T. FUKETA, “Behavior of Pre-hydrided Zircaloy-4 Cladding under Simulated LOCA conditions,” J. Nucl. Sci. Technol., 42[2] (2005) 209-218.
• F. NAGASE and T. FUKETA, “Fracture Behavior of Irradiated Zircaloy-4 Cladding under Simulated LOCA Conditions,” J. Nucl. Sci. Technol., 43[9] (2006) 1114-1119.
• F. NAGASE, T. CHUTO, and T. FUKETA, “Behavior of High Burn-up Fuel Cladding under LOCA Conditions,” J. Nucl. Sci. Technol., 46[7] (2009) 763−769.
• F. NAGASE, T. CHUTO, T. FUKEATA “Ring-compression ductility of high burn-up fuel cladding after exposure to simulated LOCA conditions”, J. Nucl. Sci. & Tech., 48[11] (2011) 1369–1376.