International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010 Incorporated administrative agency Japan Nuclear

International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010

Incorporated administrative agency Japan Nuclear Energy Safety Organization

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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)



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　　　　　　　　　　　　　　 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 　



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1. Current status of spent fuel interim storage in Japan and regulation process



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　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



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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



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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



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2. Research to investigate fundamental safety functions of Transport/Storage Cask



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　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)



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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



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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)



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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)



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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



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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



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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)




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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




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3. Research to investigate integrity of spent fuel during storage



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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



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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.

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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



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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)



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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



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4. Safety Analysis



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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



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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



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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



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5. Ongoing and future activities



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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



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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

Documents

International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010 Incorporated administrative agency Japan Nuclear