Activities Related to Safety Regulations of Spent Fuel Interim Storage at Japan Nuclear Energy Safety Organization (JNES

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)

2. Contents • Current status of spent fuel interim storage in Japan and Regulation Process • Research to investigate fundamental safety functions of transport/storage caskfor long term storage • Research to investigate integrity of spent fuel during storage • Safety Analysis • Ongoing and future activities • Summary

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

4. Current Status of Spent Fuel Interim Storage in Japan Approval of license application : May 2010 Design and Construction Methods Welding Inspection Pre-Service Inspection Commencement of operation : July 2012 Source： HP of Recyclable-Fuel Storage Company Commencement of operation : 2016 FY Source： HP of Chubu Electric Power Site investigation 2009 - 2011 Source： HPof Kyushu Electric Power

5. Flow of Nuclear Safety Regulation and Role of JNES(1/2) Technical Support :Data for fundamental safety function Safety Review Independent analysis to validate safety assessment by applicant Approval of Design and Construction Methods Technical Support : Preparation of technical Criteria Welding inspection Preparation of inspection procedure Support Order

6. Flow of Nuclear Safety Regulation and Role of JNES(2/2) Preparation of Inspection Procedure Pre-Service Inspection Inspection (in part) Approval of Operational Safety Program Continuous accumulation of degradation phenomena Operational Safety Inspection Preparation of Inspection Procedure Annual Inspection Inspection (in part) Transportation method confirmation Confirmation of consignment Support Order

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

8. Scope of Research and Examination for Fundamental Safety Function of Cask Material property changes with time during long-term storage and safety function Material and Component Safety 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 concrete cask canister ・Stress corrosion cracking of canister materials (CRIEPI) CRIEPI:Central Research Institute of Electric Power Industry

9. Possible Degradation Phenomena of Metal Cask Component *1) NS: No Significance, 　*2) Mainly due to degraded inner atmosphere

10. Test for degradation of metal cask components Material Property(1/2)

11. Test for degradation of metal cask components Material Property(2/2)

12. Test for degradation of metal cask components Safety Functions

13. Test results Material Property of Borated Al alloy for Basket 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 results Annealing 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. Comparison of proof strength (at 250 degree C) 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

14. 130 C (Non-irradiated) 150 C (Non-irradiated) 170 C (Non-irradiated) 130 C (irradiated) 150 C (irradiated) 170 C (irradiated) Weight Loss (%) Actual condition estimated 1.55*10-3 * LMP－25.3 ( C = 35 ) Degradation of Epoxy Resin (in closed system with forced ventilation) Test results Neutron 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 results Relations 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. 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

15. Results of “Degradation tests for metal gasket“ Leak Rate (Pa･m3/s) Horizontal Drop (1) Horizontal Drop (2) Vertical Drop Corner Drop Radial Displacement (mm) Results of 9m drop tests and thermal tests for lid containment behavior and seal performance Drop Tests using Full Size Cask 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. Horizontal Drop with Full size CASK Results Leak 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. 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

16. Research to investigate integrity of spent fuel during storage

17. Background and JNES Test Plan for Evaluation of Fuel Integrity Technical Issues to be Evaluated Technical Requirements in Japan 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

18. Spent Fuel Cladding Integrity Test -Summary To develop the data for safety regulation, following mechanical property tests were carried out from 2000 to 2008, using BWR and PWR fuel cladding 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

19. 100 Zry-4 cladding strain tertiary eTh secondary time primary Unirrad. eTh : Threshold strain to tertiary creep 10 Threshold strain to tertiary creep (%) Irrad. 1 : tertiary creep was not observed in the test 0.1 10-7 10-6 10-5 10-4 10-3 Secondary creep rate (1/h) Thermal Creep Test The threshold strain of transition to tertiary creep was larger than 1% for irradiated cladding, 10 % for unirradiated cladding.

20. 10-4 420ºC 390ºC 10-5 e :Secondary creep rate High stress region 360ºC 330ºC 10-6 Low stress region 10-7 10-8 BWR 50GWd/t type 10-9 10-2 10-4 10-3 s/E Thermal Creep Test Stress dependency of secondary creep rate Creep equation (1) e :Creep strain e sp: Saturated primary creep strain t : Time ：Secondary creep rate in the low stress region ：Secondary creep rate in the high stress region (s: Hoop stress, E :Young’s modulus) Creep rate was measured as parameters of stress and temperature using irradiated and unirradiated fuel cladding tubes. As results of creep test, it was shown that stress dependency of secondary creep rate was different by stress regions, cladding types and irradiation. Creep strain was expressed by equation(1) for BWR and PWR respectively.

21. Hydride Effect Evaluation Test Mechanical Property after Hydride Reorientation for PWR Zry-4 Crosshead displacement ratio : index of ductility 48GWd/t type HRT 300ºC, 115MPa, 30℃/h Hoop stress during hydride reorientation treatment (MPa) Ring compression test was carried out to evaluate the effect of temperature and stress on degradation of mechanical property. Limit condition was determined by relative comparison with the value of as-irradiated fuel cladding tube.

22. Safety Analysis

23. Safety Analysis 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. • Input Date • ・Open to the public data • ・Offered data Confirmed to satisfy the criteria • 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 Safety analysis Confirmed that the applicant analysis is appropriate • Check on applicant data • Check on applicant analysis condition

24. Analytical Code for Independent analysis

25. Temp. ℃ 40 35 29 Temperature profile for postulated storage building calculated by FLUENT coupling with SFOKS code ●Importance of radiation heat transmission Outlet ■Contributes to the except heat of the cask Effect of decreasing cask surface temperature at about maximum 20℃ compared with case only of cooling by convection of air. Intake duct ■Heating of concrete The radiation from the barrel is received, and the temperature rises up to about the height 60℃. Metal cask Concrete floor ●Heat radiation analysis code ■S-FOKS code Concrete ceiling Metal cask Calculated by FLUENT coupling with S-FOKS code Temperature profile calculated by FLUENT

26. Ongoing and future activities

27. Ongoing and Future Activities • 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. • Preparation of technical criteria for design and construction method approval • Continuous improvement of safety analysis code and method • Continuous accumulation of long term behavior of cask and spent fuel • Demonstration test program for long term storage of PWR spent fuel by utilities

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