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International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
1
Activities Related to Safety Regulations of Spent Fuel Interim Storage at Japan Nuclear
Energy Safety Organization (JNES)
M.Kato, R.Minami and K.Maruoka
Japan Nuclear Energy Safety Organization (JNES)
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
2
Contents
1. Current status of spent fuel interim storage in Japan and Regulation Process
2. Research to investigate fundamental safety functions of transport/storage cask for long term storage
3. Research to investigate integrity of spent fuel during storage
4. Safety Analysis
5. Ongoing and future activities
6. Summary
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
3
1. Current status of spent fuel interim storage in Japan and regulation process
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
4
Current Status of Spent Fuel Interim Storage in Japan Project
Mutsu ISFSF (AFR) max. 3,000 tU
Chubu Electric Power Hamaoka NPP :max. 700 tU Metal Cask
Kyushu Electric Power ISFSF
Approval of license application : May 2010 Design and Construction Methods
Welding Inspection Pre-Service Inspection
Source : HP of Recyclable-Fuel Storage Company
Source : HP of Chubu Electric Power
Source : HP of Kyushu Electric Power
Site investigation 2009 - 2011
Commencement of operation : 2016 FY
Commencement of operation : July 2012
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
5
Flow of Nuclear Safety Regulation and Role of JNES(1/2)
Stage NISA JNESPlanning and Design Stage
Construction Stage
Technical Support :Data for fundamental safety function
Technical Support :Data for fundamental safety function
Independent analysis to validate safety assessment by applicant
Independent analysis to validate safety assessment by applicant
Technical Support : Preparation of technical Criteria
Technical Support : Preparation of technical Criteria
SupportOrder
Safety ReviewSafety Review
Approval of Design and Construction Methods
Approval of Design and Construction Methods
Welding inspectionPreparation of inspection procedure
Welding inspectionPreparation of inspection procedure
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
6
Support Order
Flow of Nuclear Safety Regulation and Role of JNES(2/2)
Stage NISA JNESOperation Stage
Approval of Operational Safety Program
Approval of Operational Safety Program
Inspection (in part) Inspection (in part)
Preparation of Inspection Procedure Preparation of Inspection Procedure
Pre-Service InspectionPre-Service Inspection
Annual InspectionAnnual Inspection
Operational Safety Inspection
Operational Safety Inspection
Continuous accumulation of degradation phenomenaContinuous accumulation of degradation phenomena
Preparation of Inspection Procedure Preparation of Inspection Procedure
Inspection (in part) Inspection (in part)
Confirmation of consignmentConfirmation of consignment Transportation method confirmation Transportation method confirmation
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
7
2. Research to investigate fundamental safety functions of Transport/Storage Cask
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
8
Scope of Research and Examination for Fundamental Safety Function of Cask
Material property changes with time during long-term storage and safety functionMaterial property changes with time during long-term storage and safety function
Material and Component Material and Component Safety FunctionsSafety Functions◇Test for degradation of
metal cask components◇Examination of containment
mechanisms after long-term storage
・ Drop Test(9m drop) ・ Thermal Test(fire
condition)
◇Test for degradation of concreteconcrete cask canister
CRIEPI:Central Research Institute of Electric Power Industry
・ Stress corrosion cracking of canister materials (CRIEPI)
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
9
Heat Radiation Atmosphere
Cask body, Lid(Carbon steel, Stainless steal)
NS(*1) NS(*1) Corrosion, SCC (*2)
Basket(Borated Aluminum alloy)
Overaging, Creep NS(*1) Corrosion, SCC (*2)
Neutron shielding(Resin, Propylene glycol(PG)-water)
Composition change
Composition change
-
Seal boundary (Metal gasket)
relaxation NS(*1) Corrosion, SCC (*2)
*1) NS: No Significance, *2) Mainly due to degraded inner atmosphere
Possible Degradation Phenomena of Metal Cask Component
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
10
Test for degradation of metal cask components
Tests Purpose Main Results
Material of Cask Body / Lid (carbon steel, stainless steel, aluminum)
Confirmation of corrosion characteristic of cask material due to cask internal atmosphere deterioration
In Iodine atmosphere assuming 1 % fuel failure , SCC did not occur and corrosion is a little.
Material of basket (borated aluminum alloy)
Confirmation of long-term material strength characteristic of basket material.
Mechanical, thermal properties etc. were obtained when thermal ageing or additional creep deformation was applied. No important change was observed.
Material Property(1/2)
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
11
Test for degradation of metal cask components
Tests Purpose Main Results
Neutron shielding materials (epoxy resin, silicon resin, propylene glycol water)
Confirmation of long term shielding performance
Influence of radiation is negligible. Degradation rate of both resins caused by thermal ageing was obtained.
Metal gasket ( type: single or double, material: high nickel alloy for spring, aluminum for outer jacket)
Confirmation of relaxation change due to thermal aging
An amount of relaxation due to thermal aging was obtained. Evaluation method of leak rate from lid with relaxed metal gasket were proposed, based on experimentally obtained leak rate trend data for displacement of lid.
Material Property(2/2)
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
12
Test for degradation of metal cask components
Tests Purpose Main Results
Lid seal performance after 9m drop
In drop accident in transport after long-term storage, confirmation of integrity of confinement.
Leak rate from lid was less than 1x10-5 Pa·m3/s. Evaluation method for leak rate from lid with relaxed metal gasket at drop event were verified.Applicability of DYNA-3D code to estimate displacement of lid were verified.
Lid seal performance during fire condition (thermal test, 30 minutes 800 ºC )
In fire accident in transport after long-term storage, confirmation of integrity of confinement.
Maintaining containment safety of lid with relaxed metal gasket during fire event were confirmed.
Safety Functions
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
13
Test resultsMaterial Property of Borated Al alloy for Basket
0
20
40
60
80
100
120
0.2%
Pro
of S
tres
s (M
Pa)
初期材過時効材過時効+クリーフ 材゚
A50
52-H
34
6N01
(5%
B4C
)
A63
51 -T5
(1%
B)
A30
04-H
112
(1%
B)
InitialAnnealedAnnealed+Creeping
Comparison of proof strength (at 250 degree C)
Subjects (Metals) JIS H4080 A5052 H34 (No boron) 5wt%B4C Borated Aluminum Alloy (Base: JIS H4100 A6N01) 1wt% over Borated Aluminum Alloy (Base: ASTM A6351-T5) 1wt% Borated Aluminum Alloy (Base: ASTM A3004-H112)Annealing Condition: (200 C, 250 C), (1,000hrs, 3,000hrs, 10,000hrs)Testing Temperature: Tensile Test (200 C, 250 C), Impact Test (-20 C), Hardness (RT), Micro Structure Modulus, Thermal Conductivity & Specific Heat, Coefficient of Liner Expansion (RT, 100 C, 200 C, 250 C)Mechanical Properties for Annealed and Creeping Metal Annealing Condition: 250C, 1,000hours Creep Deformation: about 0.1 % – about 1.0% (Max.) Test Temperature (Tensile): 250 C
Test resultsAnnealing made strengths lower. Further, these strengths were almost same if additional creep deformation was provided.Absorbing energy at impact test were almost same or more than initial.There was no important change for micro structure and the other properties.
Source : Interface issues between storage safety and post-storage transport safety“Technical Meeting on Potential Interface Issues in Spent Fuel Management”, 3–6 Nov 2009
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
14
Test resultsNeutron Shielding Materials
For Epoxy resin & Silicon resin; irradiation tests of neutron or gamma radiation, heating tests after irradiation, heating tests etc.Degradation condiution : 130 C to 170 C, Max. heating time: 15,000 hrs.
Test resultsRelations of weight loss and LMP (Larson ・Muller ・ Parameter) LMP=T ( C + log t ) T: absolute temperature of heating (K), C: constant, t: heating time (hour)Weight loss was estimated to occur by release of oxide products of low molecular weight from base materials and H2O due to dehydrate reaction of tri-hydrate-alumina.Heating was dominant for weight loss.There was no synergistic effect of heating and irradiation.
0
1
2
3
4
5
6
15000 16000 17000 18000 19000 20000 21000劣化パラメータ
(%)
重量
減損
130℃非照射150℃非照射170℃非照射130℃照射150℃照射170℃照射95%信頼下限95%信頼上限
130 C (Non-irradiated)150 C (Non-irradiated)170 C (Non-irradiated)130 C (irradiated)150 C (irradiated)170 C (irradiated)
1.55*10-3 * LMP - 25.3 ( C = 35 )
Actual condition
estimated
Wei
ght L
oss
(%)
Degradation of Epoxy Resin(in closed system with forced ventilation)
Source : Interface issues between storage safety and post-storage transport safety“Technical Meeting on Potential Interface Issues in Spent Fuel Management”, 3–6 Nov 2009
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
15
Results of 9m drop tests and thermal tests for lid containment behavior and seal performance
CASK Position Horizontal Drop (1) & (2) Vertical Drop with Lid Down Corner Drop with Lid DownDrop * For Horizontal (1), metal gaskets were prepared thermal degradation. LMP=7400 was achieved.
Drop Tests using Full Size Cask
ResultsLeak rate of the secondary lid containment system with relaxed metal gasket was estimated lower than10-4 Pa ・ m3/sec on the drop of each position.Metal gasket elementary test results, radial direction of dynamic, agreed to full size cask drop.Lid behavior in drop event was simulated well by DYNA-3D code.
1.E-11
1.E-10
1.E-09
1.E-08
1.E-07
1.E-06
1.E-05
1.E-04
0.0 1.0 2.0 3.0 4.0 5.0 6.0(mm)横ずれ量
(Pa
m漏
えい
率・
3/s
ec)
9G34(φ 10 Ra=0.19 =7℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G33(φ 10 Ra=0.15 =5℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G35(φ 10 Ra=0.30 =11℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G36(φ 10 Ra=0.15 =6℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G48(φ 10 Ra=0.77 =7℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G47(φ 10 Ra=0.65 =9℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G43(φ 10 Ra=0.79 =27℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G44(φ 10 Ra=1.08 =30℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G45(φ 10 Ra=2.90 =28℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G46(φ 10 Ra=3.12 =29℃ =0 700mm/ sec B )、 、温度 、速度 ~ 、 社9G41(φ 10 Ra=0.22 =28℃ =0 500mm/ sec B )、 、温度 、速度 ~ 、 社9G42(φ 10 Ra=0.22 =30℃ =0 500mm/ sec B )、 、温度 、速度 ~ 、 社B-2(φ 10 Ra=0.30 =30℃ =0 700mm/ sec A )、 、温度 、速度 ~ 、 社B-6(φ 10 Ra=0.22 =30℃ =0 700mm/ sec A )、 、温度 、速度 ~ 、 社
1水平落下2水平落下
上部垂直落下上部コーナ落下
Results of “Degradation tests for metal gasket“
Horizontal Drop (1)Horizontal Drop (2)Vertical DropCorner Drop
Le
ak
Ra
te (
Pa・
m3 /
s)
Radial Displacement (mm)
Horizontal Drop with Full size CASK
Source : Interface issues between storage safety and post-storage transport safety“Technical Meeting on Potential Interface Issues in Spent Fuel Management”, 3–6 Nov 2009
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
16
3. Research to investigate integrity of spent fuel during storage
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
17
To prevent the failure of fuel due to cladding thermal creep
Thermal creep
To prevent the degradation of cladding mechanical properties
» Hydride reorientation
» Irradiation hardening recovery
Technical Issues to be EvaluatedTechnical Requirements in Japan
Background and JNES Test Plan for Evaluation of Background and JNES Test Plan for Evaluation of Fuel IntegrityFuel Integrity
2000 2001 2002 2003 2004 2005 2006 2007 2008
PWR48GWd/t, BWR50GWd/t PWR55GWd/t, BWR55GWd/t
PWR48GWd/t, BWR50GWd/t
Hydride Effects Evaluation Test PWR48GWd/t, 55GWd/t
» Hydride Reorientation Test BWR40GWd/t, 50GWd/t, 55GWd/t
» Mechanical Property TestPWR48GWd/t, BWR50GWd/t (330-420ºC) (<330ºC)Irradiation Hardening Recovery Test
Item FY
Survey and Planning
Creep TestCreep Test
Creep Rupture Test
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
18
Spent Fuel Cladding Integrity Test -Spent Fuel Cladding Integrity Test - SummarySummary
To develop the data for safety regulation, following mechanical propertytests were carried out from 2000 to 2008, using BWR and PWR fuelcladding tubes irradiated in commercial power reactors in Japan.
(1) Thermal creep test, creep rupture test » Threshold strain of transition to tertiary creep region is larger than 1% for irradiated cladding. » Creep equations were obtained for BWR and PWR claddings.
(2) Hydride reorientation and mechanical properties test » Based on the experimental results, limit values of temperature and stress in the dry storage were determined.
28
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
19
Thermal Creep TestThermal Creep Test
The threshold strain of transition to tertiary creep was larger than1% for irradiated cladding, 10 % for unirradiated cladding.
0.1
1
10
100
1E- 7 1E- 6 1E- 5 1E- 4 1E- 3
Unirradiated cladding 360℃Unirradiated cladding 390℃Unirradiated cladding420℃Unirradiated cladding 420℃Irradiated cladding 390℃Irradiated cladding 420℃Irradiated cladding 360℃Iradiated cladding 390℃Irradiated cladding 420℃
Unirradiated cladding
Irradiated cladding
: tertiary creep was not observed in the test
Thr
esho
ld s
trai
n to
tert
iary
cre
ep (
%)
100
10
1
0.110-7
Secondary creep rate (1/h)
10-6 10-5 10-4 10-3
Zry-4 cladding
Unirrad.
Irrad.
primary
secondary
tertiary
time
stra
in
ThTh : Threshold strain to tertiary creep
ThTh
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
20
Thermal Creep TestThermal Creep Test
Creep rate was measured as parameters of stress and temperature using irradiated and Creep rate was measured as parameters of stress and temperature using irradiated and unirradiated fuel cladding tubes. unirradiated fuel cladding tubes. As results of creep test, it was shown that stress dependency of secondary creep rate was As results of creep test, it was shown that stress dependency of secondary creep rate was different by stress regions, cladding types and irradiation. different by stress regions, cladding types and irradiation. Creep strain was expressed by equation(1) for BWR and PWR respectively.Creep strain was expressed by equation(1) for BWR and PWR respectively.
1E-9
1E-8
1E-7
1E-6
1E-5
1E-4
1E-4 1E-3 1E-2
σ /E
420℃390℃360℃330℃420℃-Calculated value390℃-Calculated value360℃-Calculated value330℃-Calculated value
Low stressregionnsL:1.3
High stressregion
nsH:7.7
420℃
390℃
360℃
330℃
Irradiated Zry-2 cladding
Sec
onda
ry c
reep
rat
e (1
/hr)
/E
Highstressregion
Low stressregion
BWR 50GWd/t type
10-4 10-210-3
10-5
10-4
10-6
10-7
10-8
10-9
330ºC
360ºC
390ºC420ºC s
pt
::Secondary creep rateSecondary creep rate
sspp: : Saturated primary creep strainSaturated primary creep strain
::Creep strainCreep strain
t t : : TimeTime
HL : Secondary creep rate Secondary creep rate in the low stress in the low stress
regionregion: Secondary creep rate Secondary creep rate in the high stress regionin the high stress region
H
L
Stress dependency of secondary creep rateStress dependency of secondary creep rate Creep equationCreep equation
(1)
(: Hoop stress, E :Young’s modulus)
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
21
Hydride Effect Evaluation TestHydride Effect Evaluation Test
Ring compression test was carried out to evaluate the effect of temperatureand stress on degradation of mechanical property. Limit condition was determined by relative comparison with the value ofas-irradiated fuel cladding tube.
0
5
10
15
20
25
30
0 50 100 150 200
HRT Hoop Stress (MPa)
Cro
ssh
ead
Dis
pla
cem
ent
Rat
io (
%)
HRT340ºC 30ºC/hHRT300ºC 30ºC/hHRT275ºC 30ºC/h
HRT250ºC 30ºC/hAs-irradiated
As irradiated
HRT 300ºC, 115MPa, 30 /h℃48GWd/t type
Hoop stress during hydride reorientation treatment (MPa)
Crosshead displacement ratio : index of ductility
Mechanical Property after Hydride Reorientation for PWR Zry-4Mechanical Property after Hydride Reorientation for PWR Zry-4
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
22
4. Safety Analysis
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
23
Purpose : Though an independent analysis for the applicant analysis by using analytical codes and/or methods for analyzing, to confirm whether the applicant analysis results satisfy the criteria and whether the applicant analysis is appropriate.
Confirmed to satisfy the criteria
Confirmed to satisfy the criteria
Confirmed that the applicant analysis is appropriate
Confirmed that the applicant analysis is appropriate
Safety analysisSafety analysis
Input Date・ Open to the public data・ Offered data
Maintenance of analytical code and method for analyzing
• Maintenance of mode of analysis with high reliability that reflects the latest finding etc.
• Setting method of analytical model and analysis condition like how etc. to give method of dividing analytical lattice and boundary condition
• Verification analysis
• Check on applicant data• Check on applicant analysis
condition
Safety Analysis
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
24
Analytical Code for Independent analysisStorage facility Cask
Thermal analysis
◆ Fluid dynamics code FLUENT
◆ Heat radiation analysis code S-FOKS
◆ Fluid dynamics code FLUENT
Shielding Analysis
Monte carlo code for neutron and photon transportMCNP5
Monte carlo code for neutron and photon transportMCNP5
Criticality Analysis
Monte carlo code for neutron transport MVP-II
Japanese evaluated nuclear data library JENDL-3.3
Monte carlo code for neutron transport MVP-II
Japanese evaluated nuclear data library JENDL-3.3
Structural Analysis
Impact and Structural Analysis Code LS-DYNA
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
25
Temperature profile calculated by FLUENT
● Importance of radiation heat transmission■ Contributes to the except heat of the cask
■ Heating of concrete
Effect of decreasing cask surface temperature at about maximum 20 compared with case only of ℃cooling by convection of air.
The radiation from the barrel is received, and the temperature rises up to about the height 60 .℃
● Heat radiation analysis code■ S-FOKS code
Calculated by FLUENT coupling with S-FOKS code
31約
98約
31約
8.5
約17
約
受入エリア
Concrete floorMetal cask
Intake duct
Outlet
Temp. ℃40
29
35Metal cask
Concrete ceiling
Temperature profile for postulated storage building calculated by FLUENT coupling with SFOKS code
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
26
5. Ongoing and future activities
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
27
Ongoing and Future Activities
1. Preparation of welding inspection procedure of canister (Corrosion resistance stainless steel )• Additional material properties were measured.• Applicability of multi-layer PT and UT inspections for those
materials is under investigation.
2. Preparation of technical criteria for design and construction method approval
3. Continuous improvement of safety analysis code and method
4. Continuous accumulation of long term behavior of cask and spent fuel• Demonstration test program for long term storage of PWR spent
fuel by utilities
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
Incorporated administrative agency Japan Nuclear Energy Safety Organization
28
Summary
Activities related to safety regulations of spent fuel interim storage at Japan Nuclear Energy Safety Organization is as follows.Past:
• Fundamental safety function of metal cask during long term storage.
• Seal performance under accident• Integrity of spent fuel during long term storage• Safety analysis code
Future:• Support preparing criteria in regulations at the subsequent
stage• Continuous improvement of safety analysis codes• Continuous accumulation of long term behavior of cask and
spent fuel
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