1. 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)

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 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

3. 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

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 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

5. 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) StageNISAJNES Planning and Design Stage Construction Stage Technical Support :Data for fundamental safety function Independent analysis to validate safety assessment by applicant Technical Support : Preparation of technical Criteria Technical Support : Preparation of technical Criteria Support Order Safety Review Approval of Design and Construction Methods Welding inspection Preparation of inspection procedure Welding inspection Preparation of inspection procedure

6. 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) StageNISAJNES Operation Stage Approval of Operational Safety Program Inspection (in part) Preparation of Inspection Procedure Pre-Service Inspection Annual Inspection Operational Safety Inspection Continuous accumulation of degradation phenomena Preparation of Inspection Procedure Inspection (in part) Confirmation of consignment Transportation method confirmation

7. 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

8. 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 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) concrete ◇ Test for degradation of concrete cask canister CRIEPI:Central Research Institute of Electric Power Industry ・ Stress corrosion cracking of canister materials (CRIEPI)

9. 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 HeatRadiationAtmosphere Cask body, Lid (Carbon steel, Stainless steal) NS(*1) Corrosion, SCC (*2) Basket (Borated Aluminum alloy) Overaging, CreepNS(*1) Corrosion, SCC (*2) Neutron shielding (Resin, Propylene glycol(PG)-water) Composition change - Seal boundary (Metal gasket) relaxationNS(*1)Corrosion, SCC (*2) *1) NS: No Significance, *2) Mainly due to degraded inner atmosphere Possible Degradation Phenomena of Metal Cask Component

10. 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 TestsPurposeMain 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)

11. 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 TestsPurposeMain 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)

12. 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 TestsPurposeMain 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·m 3 /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

13. 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 results Material Property of Borated Al alloy for Basket 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 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. 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. 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 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. 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 Weight Loss (%) 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

15. 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 Results Leak rate of the secondary lid containment system with relaxed metal gasket was estimated lower than10 -4 Pa ･ m 3 /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. Results of “Degradation tests for metal gasket“ Horizontal Drop (1) Horizontal Drop (2) Vertical Drop Corner Drop Leak Rate (Pa ･ m 3 /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

16. 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

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

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

20. 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 Test 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.  /E High stress region Low stress region BWR 50GWd/t type 10 -4 10 -2 10 -3 10 -5 10 -4 10 -6 10 -7 10 -8 10 -9 330ºC 360ºC 390ºC 420ºC  :  Secondary creep rate  s p : Saturated primary creep strain  s p : Saturated primary creep strain  :  Creep strain t : Time Secondary creep rate ： Secondary creep rate in the low stress region in the low stress region Secondary creep rate ： Secondary creep rate in the high stress region in the high stress region Stress dependency of secondary creep rate Creep equation (1) (  : Hoop stress, E :Young’s modulus)

21. 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 Test 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. 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-4

22. 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

23. 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 that the applicant analysis is appropriate Safety 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

24. 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 analysis Storage facilityCask 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 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

25. 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 Concrete floor Metal cask Intake duct Outlet Temp. ℃ 40 29 35 Metal cask Concrete ceiling T emperature profile for postulated storage building calculated by FLUENT coupling with SFOKS code

26. 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

27. 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

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 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