Text

Enclosure 4-1, Comparison Table of the 2019 Safety Evaluation Report (SAR) (J/61/B(U)F-96) and the 2023 SAR (J/2045/B(U)F) for the Model No. JRC-80Y-20T (Non-Proprietary)
	ML23115A082
Person / Time
Site:	07103035
Issue date:	04/03/2023
From:	Boyle R, Neely R; Edlow International Co
To:	; Division of Fuel Management
Shared Package
ML23115A059	List: ML23115A074; ML23115A075; ML23115A076; ML23115A077; ML23115A078; ML23115A081; ML23115A082; ML23115A084; ML23115A086;
References
	CAC 001794, EPID L-2023-DOT-0006
	Download: ML23115A082 (134)
	v • d • e

Comparison Table of SAR for Type JRC-80Y-20T before after note SAFETY ANALYSIS REPORT FOR JRC-80Y-20T JAPAN ATOMIC ENERGY AGENCY SAFETY ANALYSIS REPORT FOR JRC-80Y-20T JAPAN ATOMIC ENERGY AGENCY

Comparison Table of SAR for Type JRC-80Y-20T before after note C O N T E N T S Chapter I : Package description A. Introduction ********************************************************************************** (I)-1 B. Type of package ***************************************************************************** (I)-2 C. Package description-packaging ******************************************************** (I)-5 D. Contents of package *********************************************************************** (I)-41 Chapter II : Safety analyses **************************************************************** (II)-1 A. Structural analysis ************************************************************************* (II)-A-1 A.1 Structural design*********************************************************************** (II)-A-1 A.1.1 General description ************************************************************** (II)-A-1 A.1.2 Design ******************************************************************************** (II)-A-2 A.2 Weight and center of gravity ******************************************************* (II)-A-28 A.3 Mechanical property of material ************************************************** (II)-A-29 A.4 Standard for package ***************************************************************** (II)-A-37 A.4.1 Chemical and electrical reactions ******************************************* (II)-A-37 A.4.2 Low-temperature strength **************************************************** (II)-A-37 A.4.3 Containment system ************************************************************ (II)-A-50 A.4.4 Lifting device ********************************************************************** (II)-A-50 A.4.5 Tie-down device (influence of tie-down device upon package) ****** (II)-A-87 A.4.6 Pressure ***************************************************************************** (II)-A-96 A.4.7 Vibration **************************************************************************** (II)-A-97 A.5 Normal conditions of transport **************************************************** (II)-A-99 A.5.1 Thermal test *********************************************************************** (II)-A-99 A.5.1.1 Summary of pressure and temperature ***************************** (II)-A-99 A.5.1.2 Thermal expansion ********************************************************* (II)-A-103 A.5.1.3 Stress calculation *********************************************************** (II)-A-115 A.5.1.4 Comparison of allowable stress ***************************************** (II)-A-116 A.5.2 Water spray ************************************************************************ (II)-A-122 A.5.3 Free drop **************************************************************************** (II)-A-123 A.5.4 Stacking test *********************************************************************** (II)-A-123 A.5.5 Penetration ************************************************************************* (II)-A-132 A.5.6 Drop of square and edge ******************************************************** (II)-A-133 A.5.7 Summary and evaluation of the results *********************************** (II)-A-133 A.6 Accident conditions of transport *************************************************** (II)-A-136 A.6.1 Mechanical test drop test I (9 m drop) ********************************** (II)-A-137 A.6.1.1 Vertical drop ****************************************************************** (II)-A-171 A.6.1.2 Horizontal drop ************************************************************** (II)-A-214 A.6.1.3 Corner drop ******************************************************************* (II)-A-263 A.6.1.4 Oblique drop ****************************************************************** (II)-A-267 A.6.1.5 Summary of the results *************************************************** (II)-A-268 C O N T E N T S Chapter I : Package description A. Introduction ********************************************************************************** (I)-1 B. Type of nuclear fuel package *********************************************************** (I)-2 C. Package description-packaging ******************************************************** (I)-5 D. Contents of package *********************************************************************** (I)-41 Chapter II : Safety analyses **************************************************************** (II)-1 A. Structural analysis ************************************************************************ (II)-A-1 A.1 Structural design ********************************************************************** (II)-A-1 A.1.1 General description ************************************************************** (II)-A-1 A.1.2 Design ******************************************************************************* (II)-A-2 A.2 Weight and center of gravity ******************************************************* (II)-A-27 A.3 Mechanical property of material ************************************************** (II)-A-29 A.4 Standard for nuclear fuel package *********************************************** (II)-A-37 A.4.1 Chemical and electrical reactions ******************************************* (II)-A-37 A.4.2 Low-temperature strength **************************************************** (II)-A-37 A.4.3 Containment system ************************************************************ (II)-A-51 A.4.4 Lifting device ********************************************************************** (II)-A-51 A.4.5 Tie-down device (influence of tie-down device upon package) ***** (II)-A-89 A.4.6 Pressure ***************************************************************************** (II)-A-98 A.4.7 Vibration **************************************************************************** (II)-A-99 A.5 Normal conditions of transport *************************************************** (II)-A-102 A.5.1 Thermal test *********************************************************************** (II)-A-102 A.5.1.1 Summary of pressure and temperature ***************************** (II)-A-102 A.5.1.2 Thermal expansion ********************************************************* (II)-A-106 A.5.1.3 Stress calculation *********************************************************** (II)-A-119 A.5.1.4 Comparison of allowable stress **************************************** (II)-A-120 A.5.2 Water spray ************************************************************************ (II)-A-127 A.5.3 Free drop *************************************************************************** (II)-A-128 A.5.4 Stacking test *********************************************************************** (II)-A-128 A.5.5 Penetration ************************************************************************ (II)-A-137 A.5.6 Drop of square and edge ******************************************************* (II)-A-138 A.5.7 Summary and evaluation of the results ********************************** (II)-A-138 A.6 Accident conditions of transport ************************************************** (II)-A-141 A.6.1 Mechanical test drop test I (9 m drop) ********************************* (II)-A-142 A.6.1.1 Vertical drop ***************************************************************** (II)-A-176 A.6.1.2 Horizontal drop ************************************************************* (II)-A-208 A.6.1.3 Corner drop ******************************************************************* (II)-A-249 A.6.1.4 Oblique drop ***************************************************************** (II)-A-253 A.6.1.5 Summary of the results *************************************************** (II)-A-254 Refinement of description Changes of page number Refinement of description Changes of page number Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note A.6.2 Strength test drop test II (1m drop) ************************************* (II)-A-269 A.6.2.1 Summary of the result **************************************************** (II)-A-279 A.6.3 Thermal test *********************************************************************** (II)-A-280 A.6.3.1 Summary of temperature and pressure****************************** (II)-A-280 A.6.3.2 Thermal expansion ********************************************************* (II)-A-280 A.6.3.3 Comparison of allowable stress ***************************************** (II)-A-286 A.6.4 Water immersion ***************************************************************** (II)-A-301 A.6.5 Summary and evaluation of the results *********************************** (II)-A-302 A.7 Enhanced water immersion test ************************************************** (II)-A-322 A.8 Radioactive contents ****************************************************************** (II)-A-331 A.9 Package containing fissile material ********************************************** (II)-A-333 A.9.1 Package containing fissile material normal conditions of transport ************************************************************************************ (II)-A-333 A.9.2 Package containing fissile material accident conditions of transport ************************************************************************************ (II)-A-335 A.10 Appendix ******************************************************************************** (II)-A-340 A.10.1 Appendix-1 Mechanical property of material used for drop impact analysis *********** (II)--341 A.10.2 Appendix-2 Drop impact analysis of fins using dynamic analysis code LS-DYNA and comparison with impact tests ************************************************** (II)--345 A.10.3 Appendix-3 Strength analysis of lifting instrument ******************************************* (II)--361 A.10.4 Appendix-4 Compatibility with the ASME Code under design conditions ************** (II)--362 A.10.5 Appendix-5 Strength of the packaging when the external pressure equivalent to the water depth of 5,000 m acts ************************************************** (II)--378 A.10.6 Appendix-6 References ********************************************************************************** (II)--384 B. Thermal analysis ********************************************************************* (II)-B-1 Omission C. Containment analysis ********************************************************************* (II)-C-1 A.6.2 Strength test drop test II (1m drop)************************************* (II)-A-255 A.6.2.1 Summary of the result **************************************************** (II)-A-265 A.6.3 Thermal test *********************************************************************** (II)-A-266 A.6.3.1 Summary of temperature and pressure ***************************** (II)-A-266 A.6.3.2 Thermal expansion ********************************************************* (II)-A-266 A.6.3.3 Comparison of allowable stress **************************************** (II)-A-272 A.6.4 Water immersion ***************************************************************** (II)-A-287 A.6.5 Summary and evaluation of the results ********************************** (II)-A-288 A.7 Enhanced water immersion test ************************************************** (II)-A-308 A.8 Radioactive contents ****************************************************************** (II)-A-317 A.9 Package containing fissile material********************************************** (II)-A-319 A.9.1 Package containing fissile material normal conditions of transport ************************************************************************************ (II)-A-319 A.9.1.1 Water spray ****************************************************************** (II)-A-319 A.9.1.2 0.3 m drop test ************************************************************* (II)-A-319 A.9.1.3 Stacking test and 6 kg bar penetration ****************************** (II)-A-319 A.9.2 Package containing fissile material accident conditions of transport ************************************************************************************ (II)-A-321 A. 9.2.1 Normal test conditions *************************************************** (II)-A-321 A. 9.2.2 0.9 m drop test ************************************************************** (II)-A-321 A. 9.2.3 1 m drop test **************************************************************** (II)-A-324 A. 9.2.4 Thermal test***************************************************************** (II)-A-324 A. 9.2.5 0.9 m immersion *********************************************************** (II)-A-324 A. 9.2.6 Summary of damage state of nuclear fuel package ************* (II)-A-325 A.10 Appendix ******************************************************************************** (II)-A-326 A.10.1 Appendix-1 Mechanical property of material used for drop impact analysis ********** (II)--327 A.10.2 Appendix-2 Drop impact analysis of fins using dynamic analysis code LS-DYNA and comparison with impact tests ************************************************** (II)--331 A.10.3 Appendix-3 Strength analysis of lifting instrument ******************************************* (II)--347 A.10.4 Appendix-4 Compatibility with the ASME Code under design conditions ************** (II)--348 A.10.5 Appendix-5 Strength of the packaging when the external pressure equivalent to the water depth of 5,000 m acts ************************************************** (II)--364 A.10.6 Appendix-6 References ********************************************************************************** (II)--370 B. Thermal analysis ********************************************************************* (II)-B-1 Omission C. Containment analysis********************************************************************* (II)-C-1 Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Refinement of description Refinement of description Changes of page number Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note Omission D. Shielding analysis ************************************************************************** (II)-D-1 D.1 Summary ********************************************************************************* (II)-D-1 D.2 Source specifications ****************************************************************** (II)-D-3 D.2.1 Gamma source ******************************************************************** (II)-D-8 D.2.2 Neutron source ******************************************************************** (II)-D-10 D.3 Model specifications ******************************************************************* (II)-D-13 D.3.1 Analytical model ****************************************************************** (II)-D-13 D.3.2 Atomic number density in each region of shielding analytical model ************************************************************* (II)-D-20 D.4 Evaluation of shielding ************************************************************** (II)-D-23 D.5 Summary of the results and the evaluation *********************************** (II)-D-32 D.6 Appendix ********************************************************************************** (II)-D-35 D.6.1 Appendix-1 Neutron yields due to (.n) reaction ******************* (II)-D-36 D.6.2 Appendix-2 Gamma streaming calculation ************************** (II)-D-39 D.6.3 Appendix-3 References ***************************************************** (II)-D-53 E. Criticality analysis ************************************************************************* (II)-E-1 E.1 Summary ********************************************************************************* (II)-E-1 E.2 Analytical object ************************************************************************ (II)-E-3 E.2.1 Contents ***************************************************************************** (II)-E-3 E.2.2 Packaging *************************************************************************** (II)-E-7 E.2.3 Neutron poison ******************************************************************** (II)-E-7 E.3 Model specification ******************************************************************** (II)-E-8 E.3.1 Analytical model ****************************************************************** (II)-E-8 E.3.2 Atomic number density in each region of analytical model ********* (II)-E-27 E.4 Subcriticality evaluation ************************************************************ (II)-E-29 E.4.1 Analytical condition ************************************************************** (II)-E-29 E.4.2 Water immersion into the package ***************************************** (II)-E-29 E.4.3 Calculation method ************************************************************** (II)-E-30 E.4.4 Calculation results *************************************************************** (II)-E-31 E.5 Benchmark experiments ************************************************************* (II)-E-32 E.6 Summary of the results and the evaluation *********************************** (II)-E-34 E.7 Appendix ********************************************************************************** (II)-E-35 E.7.1 Appendix-1 Safety of the package under routine conditions of transport *************** (II)-E-36 E.7.2 Appendix-2 Omission D. Shielding analysis ************************************************************************* (II)-D-1 D.1 Summary ********************************************************************************* (II)-D-1 D.2 Source specifications ***************************************************************** (II)-D-3 D.2.1 Gamma source ******************************************************************** (II)-D-7 D.2.2 Neutron source ******************************************************************* (II)-D-9 D.3 Model specifications ****************************************************************** (II)-D-12 D.3.1 Analytical model ***************************************************************** (II)-D-12 D.3.2 Atomic number density in each region of shielding analytical model ************************************************************ (II)-D-19 D.4 Evaluation of shielding ************************************************************** (II)-D-22 D.5 Summary of the results and the evaluation ********************************** (II)-D-31 D.5.1 Shielding Design Features ******************************************* (II)-D-31 D.5.2 Results and evaluation ************************************************* (II)-D-31 D.6 Appendix ********************************************************************************* (II)-D-34 D.6.1 Appendix-1 Neutron yields due to (.n) reaction ****************** (II)-D-35 D.6.2 Appendix-2 Gamma streaming calculation ************************* (II)-D-38 D.6.3 Appendix-3 References***************************************************** (II)-D-52 E. Criticality analysis ************************************************************************* (II)-E-1 E.1 Summary ********************************************************************************* (II)-E-1 E.2 Analytical object *********************************************************************** (II)-E-2 E.2.1 Contents **************************************************************************** (II)-E-2 E.2.2 Packaging ************************************************************************** (II)-E-5 E.2.3 Neutron poison ******************************************************************** (II)-E-5 E.3 Model specification ******************************************************************** (II)-E-7 E.3.1 Analytical model ****************************************************************** (II)-E-7 E.3.1.1 Analytical model of nuclear fuel package in isolation ***** (II)-E-7 E.3.1.2 Analytical model of nuclear fuel package in array ********* (II)-E-7 E.3.2 Atomic number density in each region of analytical model ********* (II)-E-18 E.4 Subcriticality evaluation ************************************************************ (II)-E-20 E.4.1 Analytical condition ************************************************************* (II)-E-20 E.4.2 Water immersion into the package ***************************************** (II)-E-20 E.4.3 Calculation method ************************************************************** (II)-E-21 E.4.4 Calculation results ************************************************************** (II)-E-22 E.5 Benchmark experiments ************************************************************ (II)-E-23 E.6 Summary of the results and the evaluation *********************************** (II)-E-25 E.7 Appendix ********************************************************************************* (II)-E-26 E.7.1 Appendix-1 Safety of the package under routine conditions of transport *************** (II)-E-27 E.7.2 Appendix-2 Changes of page number Same as above Same as above Same as above Same as above Same as above Refinement of description Changes of page number Same as above Same as above Changes of page number Same as above Same as above Same as above Same as above Refinement of description Changes of page number Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note Safety of the package during the loading of the fuel elements ************* (II)-E-38 E.7.3 Appendix-3 Safety of the package under accident conditions ******************************* (II)-E-46 E.7.4 Appendix-4 Investigation of the optimum water density in the criticality evaluation *********************************************************************************** (II)-E-50 E.7.5 Appendix-5 References ********************************************************************************** (II)-E-51 Chapter III : Basic policy for quality management ********************************** (III)-A-1 A. Quality management system *********************************************************** (III)-A-1 B. Applicants responsibilities *************************************************************** (III)-B-1 C. Education and training ******************************************************************* (III)-C-1 D. Design control ******************************************************************************* (III)-D-1 E. Manufacturing order of the packaging, etc. **************************************** (III)-E-1 F. Handling and Maintenance ************************************************************** (III)-F-1 Chapter IV : Maintenance conditions of transport packaging and handling method of package **************************************************************** (IV)-A-1 A. Handling method *************************************************************************** (IV)-A-1 A.1 Loading method ************************************************************************ (IV)-A-1 A.2 Prior to shipping inspection of package ***************************************** (IV)-A-3 A.3 Unloading method ********************************************************************* (IV)-A-6 A.4 Preparation of empty packaging ************************************************** (IV)-A-7 B. Maintenance conditions ****************************************************************** (IV)-B-1 B.1 Visual inspections ********************************************************************** (IV)-B-1 B.2 Internal pressure inspections ****************************************************** (IV)-B-1 B.3 Leakage inspection ******************************************************************** (IV)-B-1 B.4 Shielding inspection ******************************************************************* (IV)-B-1 B.5 Subcriticality inspection ************************************************************* (IV)-B-2 B.6 Thermal inspection ******************************************************************** (IV)-B-2 B.7 Lifting inspection ********************************************************************** (IV)-B-2 B.8 Operational inspection *************************************************************** (IV)-B-2 Safety of the package during the loading of the fuel elements ************ (II)-E-29 E.7.3 Appendix-3 Safety of the package under accident conditions ******************************* (II)-E-35 E.7.4 Appendix-4 Investigation of the optimum water density in the criticality evaluation ********************************************************************************** (II)-E-38 E.7.5 Appendix-5 References ********************************************************************************** (II)-E-39 F. Consideration of Aging of Nuclear Fuel Package ********************************************** (II)-F-1 F.1 Aging Factors to be Considered ****************************************************************** (II)-F-1 F.2 Evaluation of Necessity of Considering Aging in Safety Analysis ******************** (II)-F-2 F.3 Aging Considerations in Safety Analysis ***************************************************** (II)-F-7 F.4 Appendix ************************************************************************************************ (II)-F-7 F.4.1 Appendix - 1 References ***************************************************************** (II)-F-7 Chapter III : Maintenance conditions of transport packaging and handling method of package **************************************************************** (III)-A-1 A. Handling method *************************************************************************** (III)-A-1 A.1 Loading method ************************************************************************ (III)-A-1 A.2 Prior to shipping inspection of nuclear fuel package *********************** (III)-A-4 A.3 Unloading method ********************************************************************* (III)-A-7 A.4 Preparation of empty packaging ************************************************** (III)-A-8 B. Maintenance conditions ****************************************************************** (III)-B-1 B.1 Visual inspections ********************************************************************* (III)-B-1 B.2 Internal pressure inspections ****************************************************** (III)-B-1 B.3 Leakage inspection ******************************************************************** (III)-B-1 B.4 Shielding inspection ****************************************************************** (III)-B-1 B.5 Subcriticality inspection ************************************************************* (III)-B-2 B.6 Thermal inspection ******************************************************************* (III)-B-2 B.7 Lifting inspection ********************************************************************** (III)-B-2 B.8 Operational inspection *************************************************************** (III)-B-2 Same as above Same as above Same as above Same as above Addition of Consideration of aging of nuclear fuel package due to the revision of the regulations Deletion due to moving to another chapter Modification for proper description due to deletion of the previous chapter Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note B.9 Maintenance of auxiliary system ************************************************* (IV)-B-2 B.10 Maintenance of valve and gasket, etc., of containment vessel ********* (IV)-B-3 B.11 Storage of transport packaging ************************************************** (IV)-B-3 B.12 Storage of records ******************************************************************** (IV)-B-3 B.13 Others ************************************************************************************ (IV)-B-3 Chapter V-I : Fabrication of packaging ************************************************** (V)-I-A-1 Chapter V-II : Modification of packaging *********************************************** (V)-II-A-1 B.9 Maintenance of auxiliary system ************************************************* (III)-B-2 B.10 Maintenance of valve and gasket, etc., of containment vessel ********* (III)-B-3 B.11 Storage of transport packaging ************************************************** (III)-B-3 B.12 Storage of records ******************************************************************** (III)-B-3 B.13 Others *********************************************************************************** (III)-B-3 Chapter IV-I : Fabrication of packaging ************************************************ (IV)-I-A-1 Chapter IV-II : Modification of packaging ********************************************** (IV)-II-A-1 Same as above Same as above Same as above Same as above Same as above Modification for proper description due to deletion of the previous chapter

Comparison Table of SAR for Type JRC-80Y-20T before after note LIST OF FIGURES

()-Fig.A.1 Package transport condition ************************************************ (I)-4

()-Fig.C.1 The external appearance of the package ******************************** (I)-7

()-Fig.C.2 The general view of the packaging ************************************** (I)-8

()-Fig.C.3 Tie down device *************************************************************** (I)-9

()-Fig.C.4 The containment boundary of the packaging ************************* (I)-10

()-Fig.C.5 The sectional view of the packaging ************************************ (I)-13

()-Fig.C.6 The sectional view of lateral fin ****************************************** (I)-14

()-Fig.C.7 Bottom fin and base plate *************************************************** (I)-15

()-Fig.C.8 Configuration of bottom fin ************************************************ (I)-16

()-Fig.C.9 The sectional view of vent and drain valves ************************** (I)-17

()-Fig.C.10 The plan of vent and drain valves (without protection cover) **** (I)-18

()-Fig.C.11 Drain valve protection cover ************************************************ (I)-19

()-Fig.C.12 Body lifting lug ***************************************************************** (I)-20

()-Fig.C.13 The sectional view of the lid ************************************************ (I)-23

()-Fig.C.14 The plan of the lid ************************************************************* (I)-24

()-Fig.C.15 Configuration of top fin ****************************************************** (I)-25

()-Fig.C.16 Leak test hole and plug ***************************************************** (I)-26

()-Fig.C.17 Vent valve protection cover ************************************************* (I)-27

()-Fig.C.18 Lid lifting lug ******************************************************************* (I)-28

()-Fig.C.19 Basket for box type fuel ****************************************************** (I)-30

()-Fig.C.20 The general view of the basket for box type fuel ******************** (I)-31

()-Fig.C.21 Basket for MNU type fuel ************************************************** (I)-32

()-Fig.C.22 The general view of the basket for MNU type fuel ***************** (I)-33

()-Fig.C.23 Configuration of the adapter *********************************************** (I)-34

()-Fig.D.1 JRR-3 Standard Aluminide Type Fuel *********************************** (I)-42

()-Fig.D.2 JRR-3 Standard Silicide Type Fuel *************************************** (I)-43

()-Fig.D.3 JRR-4 Low Enrichment Silicide Type Fuel ***************************** (I)-44

()-Fig.D.4 JRR-4 High Enrichment Instrumented Fuel (HEU) **************** (I)-45

()-Fig.D.5 JRR-3 Follower Aluminide Type Fuel ************************************ (I)-46

()-Fig.D.6 JRR-3 Follower Silicide Type Fuel **************************************** (I)-47

()-Fig.D.7 JRR-3 MNU Type Fuel (Top, Middle Fuel) ***************************** (I)-48

()-Fig.D.8 JRR-3 MNU Type Fuel (Bottom Fuel) *********************************** (I)-49 (II)-Fig.A.1 Center of gravity of the package ******************************************* (II)-A-27 (II)-Fig.A.2 Temperature dependency of mechanical property of SA-182 Type F304 and SA-240 Type 304 (equivalent to SUS304) ********* (II)-A-31 (II)-Fig.A.3 Temperature dependency of mechanical property of SA-564 ****** (II)-A-32 (II)-Fig.A.4 Yield stress of A1100-H14 **************************************************** (II)-A-33 LIST OF FIGURES

()-Fig.A.1 Transport condition of nuclear fuel package ***************************** (I)-4

()-Fig.C.1 The external appearance of the nuclear fuel package ************** (I)-7

()-Fig.C.2 The general view of the packaging ************************************** (I)-8

()-Fig.C.3 Tie down device *************************************************************** (I)-9

()-Fig.C.4 The containment boundary of the packaging ************************ (I)-10

()-Fig.C.5 The sectional view of the packaging ************************************ (I)-13

()-Fig.C.6 The sectional view of lateral fin ***************************************** (I)-14

()-Fig.C.7 Bottom fin and base plate *************************************************** (I)-15

()-Fig.C.8 Configuration of bottom fin *********************************************** (I)-16

()-Fig.C.9 The sectional view of vent and drain valves ************************* (I)-17

()-Fig.C.10 The plan of vent and drain valves (without protection cover) **** (I)-18

()-Fig.C.11 Drain valve protection cover *********************************************** (I)-19

()-Fig.C.12 Body lifting lug ***************************************************************** (I)-20

()-Fig.C.13 The sectional view of the lid ************************************************ (I)-23

()-Fig.C.14 The plan of the lid ************************************************************* (I)-24

()-Fig.C.15 Configuration of top fin ****************************************************** (I)-25

()-Fig.C.16 Leak test hole and plug **************************************************** (I)-26

()-Fig.C.17 Vent valve protection cover ************************************************* (I)-27

()-Fig.C.18 Lid lifting lug ******************************************************************* (I)-28

()-Fig.C.19 Basket for box type fuel ***************************************************** (I)-30

()-Fig.C.20 The general view of the basket for box type fuel ******************** (I)-31

()-Fig.C.21 Basket for MNU type fuel ************************************************* (I)-32

()-Fig.C.22 The general view of the basket for MNU type fuel ***************** (I)-33

()-Fig.C.23 Configuration of the adapter *********************************************** (I)-34

()-Fig.D.1 JRR-3 Standard Silicide Type Fuel ************************************** (I)-42

()-Fig.D.2 JRR-3 Follower Silicide Type Fuel *************************************** (I)-43

()-Fig.D.3 JRR-3 MNU Type Fuel (Top, Middle Fuel) ***************************** (I)-44

()-Fig.D.4 JRR-3 MNU Type Fuel (Bottom Fuel) *********************************** (I)-45 (II)-Fig.A.1 Center of gravity of the nuclear fuel package ************************** (II)-A-27 (II)-Fig.A.2 Temperature dependency of mechanical property of SA-182 Type F304 and SA-240 Type 304 (equivalent to SUS304) ********* (II)-A-31 (II)-Fig.A.3 Temperature dependency of mechanical property of SA-564 ****** (II)-A-32 (II)-Fig.A.4 Yield stress of A1100-H14 *************************************************** (II)-A-33 Refinement of description Deletion of JRR-3 aluminide fuel and JRR-4 fuel Changes of drawing number and page number due to deletion of drawings Refinement of description

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Fig.A.5 High temperature strength of A 1050 ************************************ (II)-A-34 (II)-Fig.A.6 Proof stress of A6061-T6 and AG3NE ************************************ (II)-A-35 (II)-Fig.A.7 Yield stress of metallic natural uranium ******************************** (II)-A-36 (II)-Fig.A.8 Low temperature impact value of metallic materials **************** (II)-A-39 (II)-Fig.A.9 Mechanical property of Aluminum alloy 1050 due to temperature (II)-A-39 (II)-Fig.A.10 Analytical model for fracture toughness of lid bolt ******************* (II)-A-42 (II)-Fig.A.11 Thermal expansion analytical model for hollow cylinder ********* (II)-A-43 (II)-Fig.A.12 Temperature distribution of the inner and outer surface of the shell under ambient temperature -40 ************************** (II)-A-48 (II)-Fig.A.13 Temperature distribution in the basket for box type fuel under ambient temperature -40********************** (II)-A-49 (II)-Fig.A.14 Details of the body lifting lug for 2-point lifting ******************** (II)-A-52 (II)-Fig.A.15 Details of the hole of the body lifting lug for 2-point lifting ***** (II)-A-56 (II)-Fig.A.16 Load on the package lifted by one body lifting lug ***************** (II)-A-55 (II)-Fig.A.17 Details of the body lifting lug for one point lifting ***************** (II)-A-59 (II)-Fig.A.18 Details of the body lifting lug for one-point lifting ***************** (II)-A-62 (II)-Fig.A.19 State lifting the package ************************************************** (II)-A-66 (II)-Fig.A.20 Details of the lid lifting lug for 2-point lifting *********************** (II)-A-67 (II)-Fig.A.21 Details of the lid lifting lug for 2-point lifting *********************** (II)-A-70 (II)-Fig.A.22 Geometry of the lid bolt **************************************************** (II)-A-72 (II)-Fig.A.23 Lead angle of the bolt ****************************************************** (II)-A-74 (II)-Fig.A.24 Analytical model of the thread groove ********************************* (II)-A-77 (II)-Fig.A.25 State of lifting the lid by one lid lifting lug ************************** (II)-A-79 (II)-Fig.A.26 Details of the lid lifting lug for one point lifting ******************** (II)-A-80 (II)-Fig.A.27 Details of fastening parts of the tie-down device ******************* (II)-A-88 (II)-Fig.A.28 Contact area of the fin shoe ********************************************** (II)-A-90 (II)-Fig.A.29 Geometry and analytical model of the packaging supporting bottom fin ******************************************************* (II)-A-92 (II)-Fig.A.30 Shock absorbing stand of the tie-down device ********************** (II)-A-92 (II)-Fig.A.31 Geometry of the shock absorbing stand to the traveling direction (II)-A-93 (II)-Fig.A.32 Geometry of the shock absorbing stand to the lateral direction (II)-A-94 (II)-Fig.A.33 Analytical model of the bottom fin ************************************* (II)-A-95 (II)-Fig.A.34 Temperature distribution of the packaging (Normal conditions of transport) **************************************** (II)-A-100 (II)-Fig.A.35 Temperature distribution in the basket for box type fuel ******** (II)-A-101 (II)-Fig.A.36 Temperature distribution in the basket for MNU type fuel ***** (II)-A-102 (II)-Fig.A.37 Analytical model of the packaging ************************************* (II)-A-105 (II)-Fig.A.38 Mises equivalent stress contours (Normal conditions of transport) (II)-A-106 (II)-Fig.A.39 Deformation (Normal conditions of transport) ********************** (II)-A-107 (II)-Fig.A.40 Longitudinal stress contours of the lid bolt (Normal conditions of transport) **************************************** (II)-A-108 (II)-Fig.A.5 High temperature strength of A 1050 ************************************ (II)-A-34 (II)-Fig.A.6 Proof stress of A6061-T6 and AG3NE ************************************ (II)-A-35 (II)-Fig.A.7 Yield stress of metallic natural uranium ******************************** (II)-A-36 (II)-Fig.A.8 Low temperature impact value of metallic materials *************** (II)-A-39 (II)-Fig.A.9 Mechanical property of Aluminum alloy 1050 due to temperature (II)-A-39 (II)-Fig.A.10 Analytical model for fracture toughness of lid bolt ****************** (II)-A-42 (II)-Fig.A.11 Thermal expansion analytical model for hollow cylinder ********* (II)-A-43 (II)-Fig.A.12 Temperature distribution of the inner and outer surface of the shell under ambient temperature -40 ************************* (II)-A-49 (II)-Fig.A.13 Temperature distribution in the basket for box type fuel under ambient temperature -40 ********************* (II)-A-50 (II)-Fig.A.14 Details of the body lifting lug for 2-point lifting ******************* (II)-A-53 (II)-Fig.A.15 Details of the hole of the body lifting lug for 2-point lifting **** (II)-A-57 (II)-Fig.A.16 Load on the package lifted by one body lifting lug **************** (II)-A-58 (II)-Fig.A.17 Details of the body lifting lug for one point lifting **************** (II)-A-60 (II)-Fig.A.18 Details of the body lifting lug for one-point lifting **************** (II)-A-63 (II)-Fig.A.19 State lifting the package ************************************************** (II)-A-68 (II)-Fig.A.20 Details of the lid lifting lug for 2-point lifting ********************** (II)-A-69 (II)-Fig.A.21 Details of the lid lifting lug for 2-point lifting ********************** (II)-A-72 (II)-Fig.A.22 Geometry of the lid bolt *************************************************** (II)-A-74 (II)-Fig.A.23 Lead angle of the bolt ****************************************************** (II)-A-76 (II)-Fig.A.24 Analytical model of the thread groove ******************************** (II)-A-79 (II)-Fig.A.25 State of lifting the lid by one lid lifting lug ************************** (II)-A-81 (II)-Fig.A.26 Details of the lid lifting lug for one point lifting ******************* (II)-A-82 (II)-Fig.A.27 Details of fastening parts of the tie-down device ****************** (II)-A-90 (II)-Fig.A.28 Contact area of the fin shoe ********************************************** (II)-A-92 (II)-Fig.A.29 Geometry and analytical model of the packaging supporting bottom fin ****************************************************** (II)-A-94 (II)-Fig.A.30 Shock absorbing stand of the tie-down device ********************** (II)-A-94 (II)-Fig.A.31 Geometry of the shock absorbing stand to the traveling direction (II)-A-95 (II)-Fig.A.32 Geometry of the shock absorbing stand to the lateral direction (II)-A-96 (II)-Fig.A.33 Analytical model of the bottom fin ************************************* (II)-A-97 (II)-Fig.A.34 Relationship between amplification factor and vibration ratio (II)-A-101 (II)-Fig.A.35 Temperature distribution of the packaging (Normal conditions of transport) **************************************** (II)-A-103 (II)-Fig.A.36 Temperature distribution in the basket for box type fuel ******* (II)-A-104 (II)-Fig.A.37 Temperature distribution in the basket for MNU type fuel **** (II)-A-105 (II)-Fig.A.38 Analytical model of the packaging ************************************* (II)-A-108 (II)-Fig.A.39 Mises equivalent stress contours (Normal conditions of transport) (II)-A-109 (II)-Fig.A.40 Deformation (Normal conditions of transport) ********************* (II)-A-110 (II)-Fig.A.41 Longitudinal stress contours of the lid bolt (Normal conditions of transport) **************************************** (II)-A-111 Changes of page number Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Addition of a drawing for aging assessment Changes of drawing and page number Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Fig.A.41 Location of stress classification lines************************************ (II)-A-117 (II)-Fig.A.42 Deformation of the contact surface of the lid and the body ******* (II)-A-120 (II)-Fig.A.43 Analytical model of the shell ********************************************** (II)-A-124 (II)-Fig.A.44 Shape and dimensions of bottom fin for supporting the packaging ************************************************** (II)-A-125 (II)-Fig.A.45 Analytical model of the bottom plate under compression ******** (II)-A-127 (II)-Fig.A.46 Contact surfaces of the body and the lid ****************************** (II)-A-131 (II)-Fig.A.47 Analytical model for 9 m top vertical drop analysis **************** (II)-A-142 (II)-Fig.A.48 Deformation of the package when the maximum displacement occurs in 9 m top vertical drop (at 6.9 ms) ********** (II)-A-143 (II)-Fig.A.49 Time history of displacement in the drop direction in 9 m top vertical drop ***************************************************** (II)-A-144 (II)-Fig.A.50 Time history of velocity in the drop direction in 9 m top vertical drop ***************************************************** (II)-A-145 (II)-Fig.A.51 Time history of deceleration in the drop direction in 9 m top vertical drop ***************************************************** (II)-A-146 (II)-Fig.A.52 Analytical model for 9 m bottom vertical drop analysis ********** (II)-A-148 (II)-Fig.A.53 Deformation of the package when the maximum displacement occurs in 9 m bottom vertical drop (at 4.9ms) ********************************** (II)-A-149 (II)-Fig.A.54 Time history of displacement of top fin in the drop direction in 9 m bottom vertical drop ************************************************* (II)-A-150 (II)-Fig.A.55 Time history of velocity in the drop direction in 9 m bottom vertical drop *********************************************** (II)-A-151 (II)-Fig.A.56 Time history of deceleration in the drop direction in 9 m bottom vertical drop *********************************************** (II)-A-152 (II)-Fig.A.57 Analytical model for 9 m horizontal drop analysis ****************** (II)-A-154 (II)-Fig.A.58 Deformation of the package when the maximum displacement occurs in 9 m horizontal drop analysis (at 15.2 ms)*************************** (II)-A-155 (II)-Fig.A.59 Time history of displacement in the drop direction in 9 m horizontal drop ******************************************************** (II)-A-156 (II)-Fig.A.60 Time history of velocity in the drop direction in 9 m horizontal drop ******************************************************** (II)-A-157 (II)-Fig.A.61 Time history of deceleration in the drop direction in 9 m horizontal drop ******************************************************** (II)-A-158 (II)-Fig.A.62 Analytical model for 9 m top corner drop analysis ****************** (II)-A-160 (II)-Fig.A.63 Deformation of the package when the maximum displacement occurs in 9 m top corner drop analysis (at 20.4 ms) *************************** (II)-A-161 (II)-Fig.A.64 Time history of displacement in the drop direction in 9 m top corner drop ****************************************************** (II)-A-162 (II)-Fig.A.65 Time history of velocity in the drop direction in 9 m top corner drop ****************************************************** (II)-A-163 (II)-Fig.A.66 Time history of deceleration in the drop direction (II)-Fig.A.42 Location of stress classification lines *********************************** (II)-A-121 (II)-Fig.A.43 Deformation of the contact surface of the lid and the body ****** (II)-A-125 (II)-Fig.A.44 A8nalytical model of the shell ******************************************** (II)-A-129 (II)-Fig.A.45 Shape and dimensions of bottom fin for supporting the packaging ************************************************* (II)-A-130 (II)-Fig.A.46 Analytical model of the bottom plate under compression ******* (II)-A-132 (II)-Fig.A.47 Contact surfaces of the body and the lid ***************************** (II)-A-136 (II)-Fig.A.48 Analytical model for 9 m top vertical drop analysis **************** (II)-A-147 (II)-Fig.A.49 Deformation of the package when the maximum displacement occurs in 9 m top vertical drop (at 6.9 ms) ********** (II)-A-148 (II)-Fig.A.50 Time history of displacement in the drop direction in 9 m top vertical drop **************************************************** (II)-A-149 (II)-Fig.A.51 Time history of velocity in the drop direction in 9 m top vertical drop **************************************************** (II)-A-150 (II)-Fig.A.52 Time history of deceleration in the drop direction in 9 m top vertical drop **************************************************** (II)-A-151 (II)-Fig.A.53 Analytical model for 9 m bottom vertical drop analysis ********* (II)-A-153 (II)-Fig.A.54 Deformation of the package when the maximum displacement occurs in 9 m bottom vertical drop (at 4.9ms) ********************************** (II)-A-154 (II)-Fig.A.55 Time history of displacement of top fin in the drop direction in 9 m bottom vertical drop ************************************************ (II)-A-155 (II)-Fig.A.56 Time history of velocity in the drop direction in 9 m bottom vertical drop *********************************************** (II)-A-156 (II)-Fig.A.57 Time history of deceleration in the drop direction in 9 m bottom vertical drop *********************************************** (II)-A-157 (II)-Fig.A.58 Analytical model for 9 m horizontal drop analysis ***************** (II)-A-159 (II)-Fig.A.59 Deformation of the package when the maximum displacement occurs in 9 m horizontal drop analysis (at 15.2 ms) ************************** (II)-A-160 (II)-Fig.A.60 Time history of displacement in the drop direction in 9 m horizontal drop ******************************************************* (II)-A-161 (II)-Fig.A.61 Time history of velocity in the drop direction in 9 m horizontal drop ******************************************************* (II)-A-162 (II)-Fig.A.62 Time history of deceleration in the drop direction in 9 m horizontal drop ******************************************************* (II)-A-163 (II)-Fig.A.63 Analytical model for 9 m top corner drop analysis ***************** (II)-A-160 (II)-Fig.A.64 Deformation of the package when the maximum displacement occurs in 9 m top corner drop analysis (at 20.4 ms) ************************** (II)-A-166 (II)-Fig.A.65 Time history of displacement in the drop direction in 9 m top corner drop ****************************************************** (II)-A-167 (II)-Fig.A.66 Time history of velocity in the drop direction in 9 m top corner drop ****************************************************** (II)-A-168 (II)-Fig.A.67 Time history of deceleration in the drop direction Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note in 9 m top corner drop ******************************************************** (II)-A-164 (II)-Fig.A.67 Analytical model for bottom corner drop analysis ******************* (II)-A-166 (II)-Fig.A.68 Deformation of the package when the maximum displacement occurs in 9 m bottom corner drop analysis (at 17.7 ms) ********************* (II)-A-167 (II)-Fig.A.69 Time history of displacement in the drop direction in 9 m bottom corner drop*************************************************** (II)-A-168 (II)-Fig.A.70 Time history of velocity in the drop direction in 9 m bottom corner drop*************************************************** (II)-A-169 (II)-Fig.A.71 Time history of deceleration in the drop direction in 9 m bottom corner drop*************************************************** (II)-A-170 (II)-Fig.A.72 Equivalent plastic strain generated in the lid flange after top vertical drop ******************************************************** (II)-A-173 (II)-Fig.A.73 Equivalent plastic strain generated in the body flange after top vertical drop ******************************************************** (II)-A-174 (II)-Fig.A.74 Time history of the axial stress in the lid bolt during 9 m top vertical drop ************************************************ (II)-A-175 (II)-Fig.A.75 Equivalent plastic strain generated in the lid flange after bottom vertical drop *************************************************** (II)-A-177 (II)-Fig.A.76 Equivalent plastic strain generated in the body flange after bottom vertical drop *************************************************** (II)-A-178 (II)-Fig.A.77 Time history of the axial stress in the lid bolt during 9 m bottom vertical drop **************************************************** (II)-A-179 (II)-Fig.A.78 Drain valve ********************************************************************* (II)-A-180 (II)-Fig.A.79 Valve main bolt **************************************************************** (II)-A-181 (II)-Fig.A.80 Valve protection cover bolt ************************************************* (II)-A-183 (II)-Fig.A.81 Analytical model of the neutron poison of the basket for box type fuel **************************************************************** (II)-A-187 (II)-Fig.A.82 Basket for MNU fuel type ************************************************** (II)-A-190 (II)-Fig.A.83 Geometry of JRR-3 MNU type fuel ************************************** (II)-A-203 (II)-Fig.A.84 Boundary condition for drop to the A-direction ********************** (II)-A-205 (II)-Fig.A.85 Stress for drop to the A-direction ***************************************** (II)-A-206 (II)-Fig.A.86 Boundary condition for drop to the B-direction ********************** (II)-A-208 (II)-Fig.A.87 Stress for drop to the B-direction **************************************** (II)-A-209 (II)-Fig.A.88 Equivalent plastic strain generated in the lid flange after horizontal drop ********************************************************* (II)-A-216 (II)-Fig.A.89 Equivalent plastic strain generated in the body flange after horizontal drop ********************************************************* (II)-A-217 (II)-Fig.A.90 Time history of axial stress in the lid bolt during 9 m horizontal drop ************************************************* (II)-A-218 (II)-Fig.A.91 Valve main bolt **************************************************************** (II)-A-220 (II)-Fig.A.92 Valve protection cover bolt ************************************************* (II)-A-222 (II)-Fig.A.93 Horizontal drop direction of the basket for box type fuel ********** (II)-A-224 in 9 m top corner drop ******************************************************* (II)-A-169 (II)-Fig.A.68 Analytical model for bottom corner drop analysis ****************** (II)-A-171 (II)-Fig.A.69 Deformation of the package when the maximum displacement occurs in 9 m bottom corner drop analysis (at 17.7 ms) ********************* (II)-A-172 (II)-Fig.A.70 Time history of displacement in the drop direction in 9 m bottom corner drop ************************************************** (II)-A-173 (II)-Fig.A.71 Time history of velocity in the drop direction in 9 m bottom corner drop ************************************************** (II)-A-174 (II)-Fig.A.72 Time history of deceleration in the drop direction in 9 m bottom corner drop ************************************************** (II)-A-175 (II)-Fig.A.73 Equivalent plastic strain generated in the lid flange after top vertical drop ******************************************************** (II)-A-178 (II)-Fig.A.74 Equivalent plastic strain generated in the body flange after top vertical drop ******************************************************** (II)-A-179 (II)-Fig.A.75 Time history of the axial stress in the lid bolt during 9 m top vertical drop *********************************************** (II)-A-180 (II)-Fig.A.76 Equivalent plastic strain generated in the lid flange after bottom vertical drop ************************************************** (II)-A-182 (II)-Fig.A.77 Equivalent plastic strain generated in the body flange after bottom vertical drop ************************************************** (II)-A-183 (II)-Fig.A.78 Time history of the axial stress in the lid bolt during 9 m bottom vertical drop **************************************************** (II)-A-184 (II)-Fig.A.79 Drain valve ********************************************************************* (II)-A-185 (II)-Fig.A.80 Valve main bolt *************************************************************** (II)-A-186 (II)-Fig.A.81 Valve protection cover bolt ************************************************* (II)-A-188 (II)-Fig.A.82 Basket for MNU fuel type************************************************** (II)-A-193 (II)-Fig.A.83 Geometry of JRR-3 MNU type fuel ************************************** (II)-A-201 (II)-Fig.A.84 Boundary condition for drop to the A-direction ********************** (II)-A-203 (II)-Fig.A.85 Stress for drop to the A-direction **************************************** (II)-A-204 (II)-Fig.A.86 Boundary condition for drop to the B-direction ********************** (II)-A-206 (II)-Fig.A.87 Stress for drop to the B-direction **************************************** (II)-A-207 (II)-Fig.A.88 Equivalent plastic strain generated in the lid flange after horizontal drop ********************************************************* (II)-A-210 (II)-Fig.A.89 Equivalent plastic strain generated in the body flange after horizontal drop ********************************************************* (II)-A-211 (II)-Fig.A.90 Time history of axial stress in the lid bolt during 9 m horizontal drop ************************************************ (II)-A-212 (II)-Fig.A.91 Valve main bolt *************************************************************** (II)-A-214 (II)-Fig.A.92 Valve protection cover bolt ************************************************* (II)-A-216 (II)-Fig.A.93 Horizontal drop direction of the basket for box type fuel ********* (II)-A-218 Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Deletion due to re-evaluation Changes of page number due to addition and deletion of drawings Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Fig.A.94 Analytical model of the basket for box type fuel at the X-direction drop******************************************************* (II)-A-227 (II)-Fig.A.95 Deformation of the basket for box type fuel when deceleration 167 g is acting ************************************************ (II)-A-228 (II)-Fig.A.96 Deformation of the basket for box type fuel after deceleration 167 g acting ******************************************** (II)-A-229 (II)-Fig.A.97 Equivalent plastic strain generated in the basket for box type fuel after deceleration 167 g acting ******************************************** (II)-A-230 (II)-Fig.A.98 Analytical model of the basket for box type fuel at the Y-direction drop ******************************************************* (II)-A-232 (II)-Fig.A.99 Result of stress analysis of the basket for box type fuel at the Y-direction drop ***************************************************** (II)-A-233 (II)-Fig.A.100 Analytical model at the weld zone of the basket for box type fuel ************************************************************** (II)-A-235 (II)-Fig.A.101 Analytical model at the weld zone in the frame of the basket for box type fuel ******************************************** (II)-A-238 (II)-Fig.A.102 Analytical model of the basket for MNU type fuel **************** (II)-A-239 (II)-Fig.A.103 Horizontal drop direction of the basket for MNU type fuel ***** (II)-A-240 (II)-Fig.A.104 Extent of the square shape pipe supported by one support plate of the basket for MNU type fuel ***************************************** (II)-A-245 (II)-Fig.A.105 Analytical model of the support plate of the basket for MNU type fuel *********************************************************** (II)-A-246 (II)-Fig.A.106 Horizontal drop direction of JRR-3 standard aluminide type fuel *********************************** (II)-A-249 (II)-Fig.A.107 Horizontal drop direction of JRR-3 standard silicide type fuel **************************************** (II)-A-251 (II)-Fig.A.108 Horizontal drop direction of JRR-3 follower aluminide type fuel ************************************* (II)-A-252 (II)-Fig.A.109 Horizontal drop direction of JRR-3 follower silicide type fuel ***************************************** (II)-A-255 (II)-Fig.A.110 Analytical model of JRR-3 MNU type fuel at the time of the horizontal drop *************************************** (II)-A-257 (II)-Fig.A.111 Stress in the Y-direction of JRR-3 MNU type fuel at the time of the horizontal drop ********* (II)-A-258 (II)-Fig.A.112 Horizontal drop direction of JRR-4 low enrichment silicide type fuel ******************************* (II)-A-260 (II)-Fig.A.113 Horizontal drop direction of JRR-4 high enrichment instrumented fuel (HEU) ***************** (II)-A-262 (II)-Fig.A.114 Equivalent plastic strain generated in the lid flange after 9 m top corner drop ************************************************** (II)-A-264 (II)-Fig.A.115 Equivalent plastic strain generated in the body flange after 9 m top corner drop ************************* (II)-A-265 (II)-Fig.A.94 Analytical model of the basket for box type fuel at the X-direction drop ****************************************************** (II)-A-221 (II)-Fig.A.95 Deformation of the basket for box type fuel when deceleration 167 g is acting ************************************************ (II)-A-222 (II)-Fig.A.96 Deformation of the basket for box type fuel after deceleration 167 g acting ******************************************** (II)-A-223 (II)-Fig.A.97 Equivalent plastic strain generated in the basket for box type fuel after deceleration 167 g acting ******************************************** (II)-A-224 (II)-Fig.A.98 Analytical model of the basket for box type fuel at the Y-direction drop ****************************************************** (II)-A-226 (II)-Fig.A.99 Result of stress analysis of the basket for box type fuel at the Y-direction drop ***************************************************** (II)-A-227 (II)-Fig.A.100 Analytical model at the weld zone of the basket for box type fuel ************************************************************** (II)-A-229 (II)-Fig.A.101 Analytical model at the weld zone in the frame of the basket for box type fuel ******************************************* (II)-A-232 (II)-Fig.A.102 Analytical model of the basket for MNU type fuel **************** (II)-A-233 (II)-Fig.A.103 Horizontal drop direction of the basket for MNU type fuel **** (II)-A-234 (II)-Fig.A.104 Extent of the square shape pipe supported by one support plate of the basket for MNU type fuel **************************************** (II)-A-239 (II)-Fig.A.105 Analytical model of the support plate of the basket for MNU type fuel *********************************************************** (II)-A-240 (II)-Fig.A.106 Horizontal drop direction of JRR-3 standard silicide type fuel *************************************** (II)-A-243 (II)-Fig.A.107 Horizontal drop direction of JRR-3 follower silicide type fuel ***************************************** (II)-A-245 (II)-Fig.A.108 Analytical model of JRR-3 MNU type fuel at the time of the horizontal drop *************************************** (II)-A-247 (II)-Fig.A.109 Stress in the Y-direction of JRR-3 MNU type fuel at the time of the horizontal drop ******** (II)-A-248 (II)-Fig.A.110 Equivalent plastic strain generated in the lid flange after 9 m top corner drop ************************************************** (II)-A-250 (II)-Fig.A.111 Equivalent plastic strain generated in the body flange after 9 m top corner drop ************************* (II)-A-251 Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Deletion of JRR-3 aluminide fuel Changes of drawing number and page number due to addition and deletion of drawings Same as above Same as above Same as above Deletion of JRR-4 Changes of drawing number and page number due to addition and deletion of drawings

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Fig.A.116 Time history of axial stress in the lid bolt during 9 m top corner drop ************************************************ (II)-A-266 (II)-Fig.A.117 Analytical model for case where the packaging directly hits the mild steel bar ***************************************************** (II)-A-270 (II)-Fig.A.118 Bending of the shell of the packaging ******************************** (II)-A-272 (II)-Fig.A.119 Analytical model for the case when the bottom plate directly hits the mild steel bar ************************ (II)-A-274 (II)-Fig.A.120 Situation for case where the valve protection cover directly hits the mild steel bar ********************************** (II)-A-276 (II)-Fig.A.121 Analytical model of the valve protection valve ********************* (II)-A-278 (II)-Fig.A.122 Temperature distribution of the packing (30 minutes after occurrence of the fire accident) ****************** (II)-A-282 (II)-Fig.A.123 Deformation (30 minutes after occurrence of the fire accident ) ****************** (II)-A-283 (II)-Fig.A.124 Equivalent plastic strain distribution (30 minutes after occurrence of the fire accident) ****************** (II)-A-284 (II)-Fig.A.125 Longitudinal stress contours of the lid bolt (30 minutes after occurrence of the fire accident) ****************** (II)-A-285 (II)-Fig.A.126 Deformation of the contact surface of the lid and the body ******************************************************** (II)-A-288 (II)-Fig.A.127 Thermal expansion analytical model of the hollow cylinder **** (II)-A-291 (II)-Fig.A.128 Temperature history of the basket for box type fuel contained JRR-3 standard aluminide type fuel *********************************** (II)-A-295 (II)-Fig.A.129 Temperature distribution model to thermal expansion analysis of the body at the time of fire accident *********************************** (II)-A-296 (II)-Fig.A.130 Temperature distribution in the basket for box type fuel (35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months after occurrence of the fire ) ********************************* (II)-A-298 (II)-Fig.A.131 Analytical model of the shell ******************************************* (II)-A-322 (II)-Fig.A.132 Analytical model of the bottom plate ********************************* (II)-A-324 (II)-Fig.A.133 The part of the O-ring and the leak tight test groove *********** (II)-A-330 (II)-Fig.A.10.1-1 Stress -strain curves *************************************************** (II)-A-344 (II)-Fig.A.10.2-1 Stress -strain curves *************************************************** (II)-A-347 (II)-Fig.A.10.2-2 Test piece form ************************************************************ (II)-A-349 (II)-Fig.A.10.2-3 Deformation of test piece ********************************************** (II)-A-350 (II)-Fig.A.10.2-4 Analytical model ********************************************************* (II)-A-351 (II)-Fig.A.10.2-5 Analytical conditions *************************************************** (II)-A-352 (II)-Fig.A.10.2-6 Deformation of analytical model 1 ********************************** (II)-A-355 (II)-Fig.A.10.2-7 Time history of displacement of the fin **************************** (II)-A-356 (II)-Fig.A.10.2-8 Time history of impact force generated on the fin ************** (II)-A-357 (II)-Fig.A.10.2-9 Deformation of analytical model 2 ********************************** (II)-A-358 (II)-Fig.A.10.2-10 Time history of displacement of the fin ************************** (II)-A-359 (II)-Fig.A.10.2-11 Time history of impact force generated on the fin ************ (II)-A-360 (II)-Fig.A.112 Time history of axial stress in the lid bolt during 9 m top corner drop *********************************************** (II)-A-252 (II)-Fig.A.113 Analytical model for case where the packaging directly hits the mild steel bar **************************************************** (II)-A-256 (II)-Fig.A.114 Bending of the shell of the packaging ******************************* (II)-A-258 (II)-Fig.A.115 Analytical model for the case when the bottom plate directly hits the mild steel bar *********************** (II)-A-260 (II)-Fig.A.116 Situation for case where the valve protection cover directly hits the mild steel bar ********************************* (II)-A-262 (II)-Fig.A.117 Analytical model of the valve protection valve ********************* (II)-A-264 (II)-Fig.A.118 Temperature distribution of the packing (30 minutes after occurrence of the fire accident) ****************** (II)-A-268 (II)-Fig.A.119 Deformation (30 minutes after occurrence of the fire accident ) ***************** (II)-A-269 (II)-Fig.A.120 Equivalent plastic strain distribution (30 minutes after occurrence of the fire accident) ****************** (II)-A-270 (II)-Fig.A.121 Longitudinal stress contours of the lid bolt (30 minutes after occurrence of the fire accident) ****************** (II)-A-271 (II)-Fig.A.122 Deformation of the contact surface of the lid and the body ******************************************************* (II)-A-274 (II)-Fig.A.123 Thermal expansion analytical model of the hollow cylinder *** (II)-A-277 (II)-Fig.A.124 Temperature history of the basket for box type fuel contained fuel element A ************************************************************* (II)-A-281 (II)-Fig.A.125 Temperature distribution model to thermal expansion analysis of the body at the time of fire accident ********************************** (II)-A-282 (II)-Fig.A.126 Temperature distribution in the basket for box type fuel (35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months after occurrence of the fire ) ******************************** (II)-A-284 (II)-Fig.A.127 Analytical model of the shell ******************************************* (II)-A-308 (II)-Fig.A.128 Analytical model of the bottom plate ******************************** (II)-A-310 (II)-Fig.A.129 The part of the O-ring and the leak tight test groove *********** (II)-A-316 (II)-Fig.A.10.1-1 Stress -strain curves *************************************************** (II)-A-330 (II)-Fig.A.10.2-1 Stress -strain curves *************************************************** (II)-A-333 (II)-Fig.A.10.2-2 Test piece form *********************************************************** (II)-A-335 (II)-Fig.A.10.2-3 Deformation of test piece ********************************************** (II)-A-336 (II)-Fig.A.10.2-4 Analytical model ********************************************************* (II)-A-337 (II)-Fig.A.10.2-5 Analytical conditions *************************************************** (II)-A-338 (II)-Fig.A.10.2-6 Deformation of analytical model 1 ********************************* (II)-A-341 (II)-Fig.A.10.2-7 Time history of displacement of the fin *************************** (II)-A-342 (II)-Fig.A.10.2-8 Time history of impact force generated on the fin ************* (II)-A-343 (II)-Fig.A.10.2-9 Deformation of analytical model 2 ********************************* (II)-A-344 (II)-Fig.A.10.2-10 Time history of displacement of the fin ************************* (II)-A-345 (II)-Fig.A.10.2-11 Time history of impact force generated on the fin ************ (II)-A-346 Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Changes of page number Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Fig.A.10.4-1 Geometry of the packaging ******************************************** (II)-A-367 (II)-Fig.A.10.4-2 Geometry of the lid ****************************************************** (II)-A-368 (II)-Fig.A.10.4-3 Geometry of the lid bolt ************************************************ (II)-A-369 (II)-Fig.A.10.4-4 Geometry of the valve disc ******************************************** (II)-A-370 (II)-Fig.A.10.4-5 Geometry of the valve main bolt ************************************ (II)-A-372 (II)-Fig.A.10.4-6 Geometry of the valve protection cover bolt ********************** (II)-A-374 (II)-Fig.A.10.5-1 Analytical model of the shell ******************************************** (II)-A-378 (II)-Fig.A.10.5-2 Analytical model of the bottom plate ********************************* (II)-A-381 (II)-Fig.B.1 The general view of the analytical model of the package containing the basket for box type fuel (In case of containing JRR-3 standard aluminide type fuel) ********************************** (II)-B-15 (II)-Fig.B.2 The longitudinal sectional view of the analytical model containing the basket for box type fuel (In case of containing JRR-3 standard aluminide type fuel) *********************************** (II)-B-16 (II)-Fig.B.3 The radial sectional view of the analytical model containing the basket for box type fuel(In case of containing JRR-3 standard aluminide type fuel) *********************************** (II)-B-17 (II)-Fig.B.4 The general view of the analytical model of the package containing the basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-B-19 (II)-Fig.B.5 The longitudinal sectional view of the analytical model containing the basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************** (II)-B-20 (II)-Fig.B.6 The radial sectional view of the analytical model containing the basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************** (II)-B-21 (II)-Fig.B.7 Temperature in the absence of solar insolation in case of containing JRR-3 standard aluminide type fuel (Longitudinal cross section) ************************************************ (II)-B-27 (II)-Fig.B.8 Temperature in the absence of solar insolation in case of containing JRR-3 standard aluminide type fuel (Radial cross section) ******* (II)-B-28 (II)-Fig.B.9 Temperature in the absence of solar insolation in case of containing JRR-3 MNU type fuel (Longitudinal cross section) **************** (II)-B-29 (II)-Fig.B.10 Temperature in the absence of solar insolation in case of containing JRR-3 MNU type fuel (Radial cross section) ************************** (II)-B-30 (II)-Fig.B.11 Temperature in solar insolation in case of containing JRR-3 standard aluminide type fuel (Longitudinal cross section) ************************************************ (II)-B-34 (II)-Fig.B.12 Temperature in solar insolation in case of containing JRR-3 standard aluminide type fuel (Radial cross section) **** (II)-B-35 (II)-Fig.B.13 Temperature in solar insolation in case of containing (II)-Fig.A.10.4-1 Geometry of the packaging ******************************************* (II)-A-353 (II)-Fig.A.10.4-2 Geometry of the lid ***************************************************** (II)-A-354 (II)-Fig.A.10.4-3 Geometry of the lid bolt *********************************************** (II)-A-355 (II)-Fig.A.10.4-4 Geometry of the valve disc ******************************************** (II)-A-356 (II)-Fig.A.10.4-5 Geometry of the valve main bolt ************************************ (II)-A-358 (II)-Fig.A.10.4-6 Geometry of the valve protection cover bolt ********************* (II)-A-360 (II)-Fig.A.10.5-1 Analytical model of the shell ******************************************* (II)-A-364 (II)-Fig.A.10.5-2 Analytical model of the bottom plate ******************************** (II)-A-367 (II)-Fig.B.1 The general view of the analytical model of the package containing the basket for box type fuel (In case of containing fuel element A) *************************************************************** (II)-B-15 (II)-Fig.B.2 The longitudinal sectional view of the analytical model containing the basket for box type fuel (In case of containing fuel element A) ****************************************************************** (II)-B-16 (II)-Fig.B.3 The radial sectional view of the analytical model containing the basket for box type fuel(In case of containing fuel element A) **************************************************************** (II)-B-17 (II)-Fig.B.4 The general view of the analytical model of the package containing the basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ****************************************************** (II)-B-19 (II)-Fig.B.5 The longitudinal sectional view of the analytical model containing the basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************** (II)-B-20 (II)-Fig.B.6 The radial sectional view of the analytical model containing the basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************** (II)-B-21 (II)-Fig.B.7 Temperature in the absence of solar insolation in case of containing fuel element A (Longitudinal cross section) *********************************************** (II)-B-27 (II)-Fig.B.8 Temperature in the absence of solar insolation in case of containing fuel element A (Radial cross section) ************************************ (II)-B-28 (II)-Fig.B.9 Temperature in the absence of solar insolation in case of containing JRR-3 MNU type fuel (Longitudinal cross section) *************** (II)-B-29 (II)-Fig.B.10 Temperature in the absence of solar insolation in case of containing JRR-3 MNU type fuel (Radial cross section) ************************** (II)-B-30 (II)-Fig.B.11 Temperature in solar insolation in case of containing fuel element A (Longitudinal cross section) *********************************************** (II)-B-34 (II)-Fig.B.12 Temperature in solar insolation in case of containing fuel element A (Radial cross section) *********************************** (II)-B-35 (II)-Fig.B.13 Temperature in solar insolation in case of containing Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Change of description Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note JRR-3 MNU type fuel (Longitudinal cross section) ***************** (II)-B-36 (II)-Fig.B.14 Temperature in solar insolation in case of JRR-3 MNU type fuel (Radial cross section)********************************************************* (II)-B-37 (II)-Fig.B.15 Points shown in temperature history figures *********************** (II)-B-50 (II)-Fig.B.16 Temperature history in case of containing JRR-3 standard aluminide type fuel ************************************ (II)-B-51 (II)-Fig.B.17 Temperature history in case of containing JRR-3 standard aluminide type fuel ************************************ (II)-B-52 (II)-Fig.B.18 Temperature history in case of containing JRR-3 standard aluminide type fuel *********************************** (II)-B-53 (II)-Fig.B.6.1 Heat transfer in the package ********************************************* (II)-B-59 (II)-Fig.B.6.2 JRR-3 standard aluminide type fuel in the basket *************** (II)-B-62 (II)-Fig.B.6.3 Direction of heat transfer in JRR-3 standard aluminide type fuel *********************************** (II)-B-64 (II)-Fig.B.6.4 Air area where convection is dominant in the basket for box type fuel ****************************************** (II)-B-68 (II)-Fig.B.6.5 Heat transfer in the package ********************************************* (II)-B-76 (II)-Fig.B.6.6 Heat transfer at the outer surface of the package ****************** (II)-B-80 (II)-Fig.B.6.7 Surface where geometrical factor is 1.0 ****************************** (II)-B-84 (II)-Fig.B.6.8 Explanation of fin and ambient air ************************************* (II)-B-85 Omission (II)-Fig.D.1 Gamma shielding analytical model with basket for box type fuel (In case of containing JRR-3 standard silicide type fuel) **************************************** (II)-D-16 (II)-Fig.D.2 Neutron shielding analytical model with basket for box type fuel (In case of containing JRR-4 low enrichment silicide type fuel) ******************************* (II)-D-17 (II)-Fig.D.3 Gamma shielding analytical model with basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-D-18 (II)-Fig.D.4 Neutron shielding analytical model with basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-D-19 (II)-Fig.D.5 Gamma dose equivalent rate [Basket for box type fuel (Axial direction)] (In case of containing JRR-3 standard silicide type fuel) **************************************** (II)-D-25 (II)-Fig.D.6 Gamma dose equivalent rate [Basket for box type fuel (Radial direction)] (In case of containing JRR-3 standard silicide type fuel) **************************************** (II)-D-26 JRR-3 MNU type fuel (Longitudinal cross section) ***************** (II)-B-36 (II)-Fig.B.14 Temperature in solar insolation in case of JRR-3 MNU type fuel (Radial cross section) ******************************************************** (II)-B-37 (II)-Fig.B.15 Points shown in temperature history figures *********************** (II)-B-50 (II)-Fig.B.16 Temperature history in case of containing fuel element A ***************************************************************** (II)-B-51 (II)-Fig.B.17 Temperature history in case of containing fuel element A ***************************************************************** (II)-B-52 (II)-Fig.B.18 Temperature history in case of containing fuel element A ***************************************************************** (II)-B-53 (II)-Fig.B.6.1 Heat transfer in the nuclear fuel package **************************** (II)-B-59 (II)-Fig.B.6.2 State of fuel element A in the fuel basket *************************** (II)-B-62 (II)-Fig.B.6.3 Direction of heat transfer in fuel element A **************************************************************** (II)-B-64 (II)-Fig.B.6.4 Air area where convection is dominant in the basket for box type fuel ***************************************** (II)-B-68 (II)-Fig.B.6.5 Heat transfer in the nuclear fuel package **************************** (II)-B-76 (II)-Fig.B.6.6 Heat transfer at the outer surface of the package ***************** (II)-B-80 (II)-Fig.B.6.7 Surface where geometrical factor is 1.0 ***************************** (II)-B-84 (II)-Fig.B.6.8 Explanation of fin and ambient air ************************************* (II)-B-85 Omission (II)-Fig.D.1 Gamma shielding analytical model with basket for box type fuel (In case of containing JRR-3 standard silicide type fuel) **************************************** (II)-D-15 (II)-Fig.D.2 Neutron shielding analytical model with basket for box type fuel (In case of containing fuel element B) ***************************************************************** (II)-D-16 (II)-Fig.D.3 Gamma shielding analytical model with basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-D-17 (II)-Fig.D.4 Neutron shielding analytical model with basket for MNU type fuel (In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-D-18 (II)-Fig.D.5 Gamma dose equivalent rate [Basket for box type fuel (Axial direction)] (In case of containing JRR-3 standard silicide type fuel) **************************************** (II)-D-24 (II)-Fig.D.6 Gamma dose equivalent rate [Basket for box type fuel (Radial direction)] (In case of containing JRR-3 standard silicide type fuel) **************************************** (II)-D-25 Same as above Same as above Same as above Same as above Same as above Same as above Same as above Changes of page number Change of description Changes of page number Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Fig.D.7 Gamma dose equivalent rate [Basket for MNU type fuel (Axial direction)](In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-D-27 (II)-Fig.D.8 Neutron dose equivalent rate [Basket for box type fuel (Axial direction)] (In case of containing JRR-4 low enrichment silicide type fuel) ******************************* (II)-D-30 (II)-Fig.D.9 Neutron dose equivalent rate [Basket for MNU type fuel (Axial direction)] (In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-D-31 (II)-Fig.D.2.1 Evaluated Location of Gamma Streaming**************************** (II)-D-40 (II)-Fig.D.2.2 Streaming analytical model for the drain hole ********************* (II)-D-42 (II)-Fig.D.2.3 Gap between the body and the lid ************************************* (II)-D-46 (II)-Fig.D.2.4 Three routes of leakage *************************************************** (II)-D-47 (II)-Fig.D.2.5 Gamma streaming analytical model for (B) ************************ (II)-D-48 (II)-Fig.D.2.6 Gamma streaming analytical model for (C) ************************ (II)-D-48 (II)-Fig.D.2.7 Gamma streaming analytical model of the vent hole ************ (II)-D-51 (II)-Fig.E.1 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Axial direction)

[In case of containing JRR-3 standard aluminide or silicide type fuel] ********************** (II)-E-11 (II)-Fig.E.2 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Cross section of basket)

[In case of containing JRR-3 standard aluminide or silicide type fuel] ********************** (II)-E-12 (II)-Fig.E.3 Cross section of JRR-3 standard aluminide type fuel ***************** (II)-E-13 (II)-Fig.E.4 Cross section of JRR-3 standard silicide type fuel ********************** (II)-E-14 (II)-Fig.E.5 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Axial direction)

[In case of containing JRR-4 low enrichment silicide type fuel] ******************************* (II)-E-15 (II)-Fig.E.6 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Cross section of basket)

[In case of containing JRR-4 low enrichment silicide type fuel] ******************************* (II)-E-16 (II)-Fig.E.7 Cross section of JRR-4 low enrichment silicide type fuel ************* (II)-E-17 (II)-Fig.E.8 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Axial direction)

[In case of containing JRR-4 high enrichment instrumented fuel (HEU)] ****************** (II)-E-18 (II)-Fig.E.9 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Cross section of basket)

(II)-Fig.D.7 Gamma dose equivalent rate [Basket for MNU type fuel (Axial direction)](In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-D-26 (II)-Fig.D.8 Neutron dose equivalent rate [Basket for box type fuel (Axial direction)] (In case of containing fuel element B) ***************************************************************** (II)-D-29 (II)-Fig.D.9 Neutron dose equivalent rate [Basket for MNU type fuel (Axial direction)] (In case of containing JRR-3 MNU type fuel) ******************************************************* (II)-D-30 (II)-Fig.D.2.1 Evaluated Location of Gamma Streaming *************************** (II)-D-39 (II)-Fig.D.2.2 Streaming analytical model for the drain hole ******************** (II)-D-41 (II)-Fig.D.2.3 Gap between the body and the lid ************************************* (II)-D-45 (II)-Fig.D.2.4 Three routes of leakage ************************************************** (II)-D-46 (II)-Fig.D.2.5 Gamma streaming analytical model for (B) ************************ (II)-D-47 (II)-Fig.D.2.6 Gamma streaming analytical model for (C) ************************ (II)-D-47 (II)-Fig.D.2.7 Gamma streaming analytical model of the vent hole *********** (II)-D-50 (II)-Fig.E.1 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Axial direction)

[In case of containing JRR-3 silicide type fuel] ***************************************************** (II)-E-10 (II)-Fig.E.2 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Cross section of basket)

[In case of containing JRR-3 silicide type fuel] ***************************************************** (II)-E-11 (II)-Fig.E.3 Cross section of JRR-3 standard silicide type fuel ********************* (II)-E-12 Same as above Change of description Changes of page number Same as above Same as above Same as above Same as above Same as above Same as above Deletion of JRR-3 aluminide fuel Changes of drawing number and page number due to addition and deletion of drawings Deletion of JRR-4

Comparison Table of SAR for Type JRC-80Y-20T before after note

[In case of containing JRR-4 high enrichment instrumented fuel (HEU)] ******************* (II)-E-19 (II)-Fig.E.10 Cross section of JRR-4 high enrichment instrumented fuel (HEU) ******************* (II)-E-20 (II)-Fig.E.11 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Axial direction)

[In case of containing JRR-3 follower aluminide or silicide type fuel] ************************ (II)-E-21 (II)-Fig.E.12 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Cross section of basket)

[In case of containing JRR-3 follower aluminide or silicide type fuel] ************************ (II)-E-22 (II)-Fig.E.13 Cross section of JRR-3 follower aluminide type fuel *********** (II)-E-23 (II)-Fig.E.14 Cross section of JRR-3 follower silicide type fuel **************** (II)-E-24 (II)-Fig.E.15 Analytical model of undamaged and damaged packages in array, containing the basket for MNU type fuel (Axial direction)

[In case of containing JRR-3 MNU type fuel] ********************** (II)-E-25 (II)-Fig.E.16 Analytical model of undamaged and damaged packages in array, containing the basket for MNU type fuel (Cross section of basket)

[In case of containing JRR-3 MNU type fuel] **************************************************** (II)-E-26 (II)-Fig.E.17 Arrangement plan of the core ***************************************** (II)-E-33 (II)-Fig.E.2.1 Analytical model in case of containing the basket for box type fuel (Axial direction)[In case of containing JRR-3 standard aluminide or silicide type fuel] ******************* (II)-E-41 (II)-Fig.E.2.2 Analytical model in case of containing the basket for box type fuel (Axial direction) [In case of containing JRR-4 low enrichment silicide type or high enrichment instrumented fuel] ********************************** (II)-E-42 (II)-Fig.E.2.3 Analytical model in case of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 follower aluminide or silicide type fuel] ******************** (II)-E-43 (II)-Fig.E.2.4 Analytical model in case of containing the basket for MNU type fuel ******************************************* (II)-E-44 (II)-Fig.E.3.1 Maximum displacement of basket for box type fuel after 9 m drop test ********************************************************* (II)-E-48 (II)-Fig.E.3.2 Analytical model of basket for box type fuel after 9 m drop test (Cross section of basket) *********************** (II)-E-49 (II)-Fig.E.4.1 Influence of water density on effective multiplication coefficient (Keff) **************************************** (II)-E-50 (II)-Fig.E.4 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Axial direction)

[In case of containing JRR-3 follower silicide type fuel] ****************************************** (II)-E-13 (II)-Fig.E.5 Analytical model of undamaged and damaged packages in array, containing the basket for box type fuel (Cross section of basket)

[In case of containing JRR-3 follower silicide type fuel] ****************************************** (II)-E-14 (II)-Fig.E.6 Cross section of JRR-3 follower silicide type fuel *************** (II)-E-15 (II)-Fig.E.7 Analytical model of undamaged and damaged packages in array, containing the basket for MNU type fuel (Axial direction)

[In case of containing JRR-3 MNU type fuel]********************** (II)-E-16 (II)-Fig.E.8 Analytical model of undamaged and damaged packages in array, containing the basket for MNU type fuel (Cross section of basket)

[In case of containing JRR-3 MNU type fuel]**************************************************** (II)-E-17 (II)-Fig.E.9 Arrangement plan of the core ***************************************** (II)-E-24 (II)-Fig.E.2.1 Analytical model in case of containing the basket for box type fuel (Axial direction)[In case of containing JRR-3 standard silicide type fuel] ************************************ (II)-E-31 (II)-Fig.E.2.2 Analytical model in case of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 follower silicide type fuel] ************************************** (II)-E-32 (II)-Fig.E.2.3 Analytical model in case of containing the basket for MNU type fuel ****************************************** (II)-E-33 (II)-Fig.E.3.1 Maximum displacement of basket for box type fuel after 9 m drop test ********************************************************* (II)-E-36 (II)-Fig.E.3.2 Analytical model of basket for box type fuel after 9 m drop test (Cross section of basket) *********************** (II)-E-37 (II)-Fig.E.4.1 Influence of water density on effective multiplication coefficient (Keff) **************************************** (II)-E-38 Deletion of JRR-3 aluminide fuel Changes of drawing number and page number due to addition and deletion of drawings Same as above Same as above Same as above Same as above Same as above Same as above Deletion of JRR-4 Deletion of JRR-3 aluminide fuel Changes of drawing number and page number due to addition and deletion of drawings Changes of page number Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note (III)-Fig.B.1 Quality assurance organization for design of the packaging, etc. ********************************************************* (III)-B-4 Deletion due to moving to another chapter

Comparison Table of SAR for Type JRC-80Y-20T before after note LIST OF TABLES (I)-Table D.1 Specification of contents....................................................... (I)-50 (I)-Table D.2 Quantities of major radionuclides (per package)................. (I)-51 (II)-Table A.1 Structural design conditions and analysis method Requirements of package1.................................... (II)-A-5 (II)-Table A.2 Weight of Package................................................................. (II)-A-28 (II)-Table A.3 Mechanical property of material......................................... (II)-A-30 (II)-Table A.4 Dissimilar materials contacting......................................... (II)-A-37 (II)-Table A.5 Standard usage for respective material of packings.......... (II)-A-40 (II)-Table A.6 Loading condition, allowable stress and safety factor of lifting lug............................................................................. (II)-A-51 (II)-Table.A.7 The maximum force applied to one bottom fin..................... (II)-A-89 (II)-Table A.8 Summary of pressure and temperature (Normal conditions of transport)........................................ (II)-A-103 (II)-Table A.9 Thermal expansion in the longitudinal direction of each basket and the packaging, safety factor and safety margin (II)-A-110 (II)-Table.A.10 Stress in Basket.................................................................. (II)-A-112 (II)-Table A.11 Maximum thermal stress generated in each basket.......... (II)-A-114 (II)-Table A.12 Thermal expansion in the radial direction of each basket and the packaging, safety factor and safety margin (II)-A-114 (II)-Table A.13 Stress of lid bolt.................................................................... (II)-A-119 (II)-Table A.14 The analysis conditions of the drop analysis........................ (II)-A-139 (II)-Table A.15 Material characteristics of JRR-3 MNU type fuel............... (II)-A-202 (II)-Table A.16 Maximum stress, safety factor and safety margin of JRR-3 MNU type fuel for the drop to A-direction.............. (II)-A-204 (II)-Table A.17 Maximum stress, safety factor and safety margin of JRR-3 MNU type fuel for the drop to B-direction.............. (II)-A-207 (II)-Table A.18 Maximum impact deceleration of drop test-I....................... (II)-A-268 (II)-Table A.19 Summary of pressure and temperature (at fire accident)... (II)-A-286 (II)-Table A.20 Result of structural analysis (Requirements of package).... (II)-A-304 (II)-Table A.21 Evaluation of the strain at the containment boundary between the lid and the shell flange under accident conditions of transport........................................................ (II)-A-321 (II)-Table A.22 Damage state of the package under normal conditions of transport for packages containing fissile material......... (II)-A-334 (II)-Table A.23 Drop posture and procedure of sequence drop tests........... (II)-A-337 (II)-Table A.24 Accumulated displacement value and impact deceleration of 9 m drop test.................................................................... (II)-A-337 (II)-Table A.25 Damage state of the package under the accident conditions of transport for package containing fissile material....................................................................... (II)-A-339 LIST OF TABLES (I)-Table D.1 Specification of contents........................................................ (I)-46 (I)-Table D.2 Quantities of major radionuclides (per package).................. (I)-47 (II)-Table A.1 Structural design conditions and analysis method Requirements of package1.................................... (II)-A-5 (II)-Table A.2 Weight of nuclear fuel package............................................. (II)-A-28 (II)-Table A.3 Mechanical property of material......................................... (II)-A-30 (II)-Table A.4 Dissimilar materials contacting.......................................... (II)-A-37 (II)-Table A.5 Standard usage for respective material of packings........... (II)-A-40 (II)-Table A.6 Loading condition, allowable stress and safety factor of lifting lug.............................................................................. (II)-A-52 (II)-Table.A.7 The maximum force applied to one bottom fin..................... (II)-A-91 (II)-Table A.8 Summary of pressure and temperature (Normal conditions of transport)......................................... (II)-A-106 (II)-Table A.9 Thermal expansion in the longitudinal direction of each basket and the packaging, safety factor and safety margin (II)-A-113 (II)-Table.A.10 Stress in Basket................................................................... (II)-A-116 (II)-Table A.11 Maximum thermal stress generated in each basket........... (II)-A-118 (II)-Table A.12 Thermal expansion in the radial direction of each basket and the packaging, safety factor and safety margin. (II)-A-118 (II)-Table A.13 Stress of lid bolt..................................................................... (II)-A-124 (II)-Table A.14 The analysis conditions of the drop analysis........................ (II)-A-144 (II)-Table A.15 Material characteristics of JRR-3 MNU type fuel................ (II)-A-200 (II)-Table A.16 Maximum stress, safety factor and safety margin of JRR-3 MNU type fuel for the drop to A-direction............... (II)-A-202 (II)-Table A.17 Maximum stress, safety factor and safety margin of JRR-3 MNU type fuel for the drop to B-direction............... (II)-A-205 (II)-Table A.18 Maximum impact deceleration of drop test-I........................ (II)-A-254 (II)-Table A.19 Summary of pressure and temperature (at fire accident).... (II)-A-272 (II)-Table A.20 Result of structural analysis (Requirements of package)..... (II)-A-290 (II)-Table A.21 Evaluation of the strain at the containment boundary between the lid and the shell flange under accident conditions of transport......................................................... (II)-A-307 (II)-Table A.22 Damage state of the package under normal conditions of transport for packages containing fissile material......... (II)-A-320 (II)-Table A.23 Drop posture and procedure of sequence drop tests........... (II)-A-323 (II)-Table A.24 Accumulated displacement value and impact deceleration of 9 m drop test.................................................................... (II)-A-323 (II)-Table A.25 Damage state of the package under the accident conditions of transport for package containing fissile material..................................................................... (II)-A-325 Changes of page number Refinement of description Changes of page number Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Table A10.2-1 Peak load and deformation value of the fin

( Test results)................................................................... (II)-A-349 (II)-Table A.10.4-1 Comparison between the allowable value and the actual size.................................................................. (II)-A-377 Omission (II)-Table D.1 Specification of fuel elements contained in the packaging.... (II)-D-6 (II)-Table D.2 Specification of fuel elements used in the shielding analysis (II)-D-7 (II)-Table D.3 Gamma source intensity......................................................... (II)-D-9 (II)-Table D.4 Neutron source intensity...................................................... (II)-D-11 (II)-Table D.5 Neutron source spectrum...................................................... (II)-D-12 (II)-Table D.6 Volume ratio of materials in source region used in gamma shielding analysis.................................................... (II)-D-21 (II)-Table D.7 Element density of each region used in gamma shielding analysis................................................................................. (II)-D-21 (II)-Table D.8 Volume ratio of materials in source region used in gamma shielding analysis................................................ (II)-D-21 (II)-Table D.9 Atomic number density in each region used in neutron shielding analysis................................................................... (II)-D-22 (II)-Table D.10 Density of materials used in shielding analysis.................. (II)-D-22 (II)-Table D.11 Conversion coefficient of unit gamma flux into air absorbed dose equivalent rate.............................................................. (II)-D-24 (II)-Table D.12 Conversion coefficient of neutron dose equivalent rate... (II)-D-29 (II)-Table D.13 Maximum dose equivalent rate in transport of the package with basket for box type fuel............................................ (II)-D-33 (II)-Table D.14 Maximum dose equivalent rate in transport of the package with basket for MNU type fuel......................................... (II)-D-34 (II)-Table D.2.1 Dose equivalent rate at each point of packaging............. (II)-D-39 (II)-Table D.2.2 Gamma flux at point P1.................................................... (II)-D-41 (II)-Table D.2.3 Gamma flux at point-P2.................................................... (II)-D-43 (II)-Table D.2.4 Gamma flux at point-P3.................................................... (II)-D-44 (II)-Table D.2.5 Linear attenuation coefficient µ......................................... (II)-D-45 (II)-Table D.2.6 Gamma flux at point-P4.................................................... (II)-D-45 (II)-Table D.2.7 Dose equivalent rate at point-P4....................................... (II)-D-46 (II)-Table D.2.8 Summary of streaming dose equivalent rate................... (II)-D-49 (II)-Table D.2.9 Gamma flux at each point................................................ (II)-D-50 (II)-Table D.2.10 Dose equivalent rate on the surface of the vent hole....... (II)-D-52 (II)-Table E.1 Fuel element contained in the packaging........................... (II)-E-5 (II)-Table E.2 Specification of fuel element used in the criticality analysis................................................................................ (II)-E-6 (II)-Table A10.2-1 Peak load and deformation value of the fin

( Test results)................................................................... (II)-A-335 (II)-Table A.10.4-1 Comparison between the allowable value and the actual size.................................................................. (II)-A-363 Omission (II)-Table D.1 Specification of fuel elements contained in the packaging.... (II)-D-5 (II)-Table D.2 Specification of fuel elements used in the shielding analysis (II)-D-6 (II)-Table D.3 Gamma source intensity......................................................... (II)-D-8 (II)-Table D.4 Neutron source intensity....................................................... (II)-D-10 (II)-Table D.5 Neutron source spectrum...................................................... (II)-D-11 (II)-Table D.6 Volume ratio of materials in source region used in gamma shielding analysis..................................................... (II)-D-20 (II)-Table D.7 Element density of each region used in gamma shielding analysis.................................................................................. (II)-D-20 (II)-Table D.8 Volume ratio of materials in source region used in gamma shielding analysis................................................. (II)-D-20 (II)-Table D.9 Atomic number density in each region used in neutron shielding analysis.................................................................... (II)-D-21 (II)-Table D.10 Density of materials used in shielding analysis................... (II)-D-21 (II)-Table D.11 Conversion coefficient of unit gamma flux into air absorbed dose equivalent rate.............................................................. (II)-D-23 (II)-Table D.12 Conversion coefficient of neutron dose equivalent rate.... (II)-D-28 (II)-Table D.13 Maximum dose equivalent rate in transport of the package with basket for box type fuel............................................. (II)-D-32 (II)-Table D.14 Maximum dose equivalent rate in transport of the package with basket for MNU type fuel.......................................... (II)-D-33 (II)-Table D.2.1 Dose equivalent rate at each point of packaging.............. (II)-D-38 (II)-Table D.2.2 Gamma flux at point P1..................................................... (II)-D-40 (II)-Table D.2.3 Gamma flux at point-P2..................................................... (II)-D-42 (II)-Table D.2.4 Gamma flux at point-P3..................................................... (II)-D-43 (II)-Table D.2.5 Linear attenuation coefficient µ.......................................... (II)-D-44 (II)-Table D.2.6 Gamma flux at point-P4..................................................... (II)-D-44 (II)-Table D.2.7 Dose equivalent rate at point-P4....................................... (II)-D-45 (II)-Table D.2.8 Summary of streaming dose equivalent rate.................... (II)-D-48 (II)-Table D.2.9 Gamma flux at each point................................................. (II)-D-49 (II)-Table D.2.10 Dose equivalent rate on the surface of the vent hole........ (II)-D-51 (II)-Table E.1 Fuel element contained in the packaging........................... (II)-E-3 (II)-Table E.2 Specification of fuel element used in the criticality analysis................................................................................ (II)-E-4 Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Table E.3 Atomic number density of each region............................... (II)-E-28 (II)-Table E.4 Calculation results................................................................ (II)-E-31 (II)-Table E.2.1 Atomic number density used in the criticality analysis in case of containing the basket for MNU type fuel........... (II)-E-45 (IV)-Table A.1 Prior to shipping inspection points....................................... (IV)-A-4 (IV)-Table B.1 Periodical inspection points................................................ (IV)-B-4 (II)-Table E.3 Atomic number density of each region................................ (II)-E-19 (II)-Table E.4 Calculation results................................................................ (II)-E-22 (II)-Table E.2.1 Atomic number density used in the criticality analysis in case of containing the basket for MNU type fuel............ (II)-E-34 (II)-Table F.1 Anticipated usage during the planned period of use............ (II)-F-1 (II)-Table F.2 Evaluation of necessity of considering aging in safety analysis (1/3)............................................................... (II)-F-3 (II)-Table F.2 Evaluation of necessity of considering aging in safety analysis (2/3)............................................................... (II)-F-5 (II)-Table F.2 Evaluation of necessity of considering aging in safety analysis (3/3)............................................................... (II)-F-6 (III)-Table A.1 Prior to shipping inspection points....................................... (III)-A-5 (III)-Table B.1 Periodical inspection points................................................. (III)-B-4 Same as above Same as above Same as above ddition of Consideration of aging of nuclear fuel package due to the revision of the regulations Modification for proper description due to deletion of the previous chapter

Comparison Table of SAR for Type JRC-80Y-20T before after note Chapter : Package description A. Introduction This packaging is dry type, and is named JRC-80Y-20T. The transportation appearance is shown in ()-Fig.A.1.

The JRC-80Y-20T packaging is used to transport spent fuels from reactors for research (JRR-3, JRR-4) of Japan Atomic Energy Agency (former Japan Atomic Energy Research Institute) to reprocessing plants in foreign countries.

A.1. Name of the packaging:

JRC-80Y-20T A.2. Type:

Type B(U) package for fissile material A.3. Allowable number of packages and allowable arrangement of packages Allowable number of packages : No restriction Allowable arrangement of packages : No restriction A.4. Transport index and criticality safety index Transport index :

Less than 5.8 Criticality safety index :

0 A.5. Maximum weight of the package:

23.2 x 103 kg (at loading the basket for box type fuel)

A.6. Size of the packaging (at body lifting lug): Diameter approx. 1.9m Height: approx. 2.1m A.7. Maximum weight of the packaging:

22.8 x 103 kg (at loading the basket for box type fuel)

A.8. Materials

1)

Main parts : Stainless steel (SA-182 Grade F304)

2)

Basket : Stainless steel (SA-182 Grade F304, SA-240 Type304), Boral plate

3)

Fin (for heat dissipation and shock absorbing) :Stainless steel (SA-240 Type 304)

A.9. Component of the packaging

1)

Body

2)

The basket (The following baskets are used for the fuel types below.)

The basket for box type fuel The basket for MNU type fuel A.10. Fuel elements contained in the packaging.

The name and number of fuel elements contained in the packaging are as follows; The basket for box type fuel 40 pieces (maximum)

1) JRR-3 standard aluminide type fuel

2) JRR-3 standard silicide type fuel

3) JRR-4 low enrichment silicide type fuel

4) JRR-4 high enrichment instrumented fuel (HEU)

5) JRR-3 follower aluminide type fuel

6) JRR-3 follower silicide type fuel

7) Fuels combined the above fuels 1) through 6)

(Herein, these fuel elements 1) through 4) are called box type fuel elements, Chapter : Package description A. Introduction This packaging is dry type, and is named JRC-80Y-20T. The transportation appearance is shown in ()-Fig.A.1.

The JRC-80Y-20T packaging is used to transport spent fuels from reactors for research (JRR-3) of Japan Atomic Energy Agency (former Japan Atomic Energy Research Institute) to reprocessing plants in foreign countries.

A.1. Name of the packaging:

JRC-80Y-20T A.2. Type:

Type B(U) package for fissile material A.3. Allowable number of packages and allowable arrangement of packages Allowable number of packages : Unlimited Allowable arrangement of packages : No restriction A.4. Transport index and criticality safety index Transport index :

Less than 5.8 Criticality safety index :

0 A.5. Maximum weight of the package:

Less than 23.2 tons (at loading the basket for box type fuel)

A.6. Size of the packaging (at body lifting lug): Diameter approx. 1.9m Height: approx. 2.1m A.7. Maximum weight of the packaging:

Less than 22.8 tons (at loading the basket for box type fuel)

A.8. Materials

1)

Main parts : Stainless steel (SA-182 Grade F304)

2)

Basket : Stainless steel (SA-182 Grade F304, SA-240 Type304), Boral plate

3)

Fin (for heat dissipation and shock absorbing) :Stainless steel (SA-240 Type 304)

A.9. Component of the packaging

1)

Body

2)

The basket (The following baskets are used for the fuel types below.)

The basket for box type fuel The basket for MNU type fuel A.10. Fuel elements contained in the packaging.

Type, number of assemblies, and number of rods to be loaded in the transport container are shown below sorted by fuel basket:

The basket for box type fuel 40 pieces (maximum)

1) JRR-3 standard silicide type fuel

2) JRR-3 follower silicide type fuel

3) Fuels combined the above fuels 1) through 2)

(Herein, the fuel element 1) is called the standard type fuel element, and the fuel Deletion of JRR-4 Modification for proper description Modification for proper description Modification for proper description Deletion of JRR-3 Aluminide Deletion of JRR-4 Changes of drawing number due to deletion of

Comparison Table of SAR for Type JRC-80Y-20T before after note and these fuel elements 5) and 6) are called follower type fuel elements)

The basket for MNU type fuel 160 pieces

1) JRR-3 MNU* type fuel Omission C. Package description - packaging Omission C.2.3 Basket (Omission)

As shown in (I)-Fig.D.1 to (I)-Fig.D.6, the size in the section of the follower fuel elements which are loaded with the basket for box type fuel is smaller than that of the standard type fuel elements. A large space (19mm when the fuel element is put aside in the lodgement.) arises between the inside of the basket lodgment and the fuel elements. Therefore, when the follower type fuel elements are loaded with the basket for type fuel, the adapter made of aluminum alloy shown in (I)-Fig.C.23 is inserted between the fuel element and the basket lodgement.

Omission C.3.4 Basket (for box type fuel)

(1) Neutron poison Boron carbide (2) Frame Stainless steel (3) Bottom plate Stainless steel (4) Compartment plate Stainless steel, Boral plate (5) Partition plate Stainless steel As supplementary part, (6) Adapter Aluminum alloy Omission C.5. Weight (a) Body (with fins) a : 13.8x103 kg (b) Lid, lid bolt (1) Lid (with fins) b1 : 6.7x103 kg (2) Lid bolt b2 : 0.2x103 kg (c) Basket (1) Basket for box type fuel c1 : 2.1x103 kg (2) Basket for MNU type fuel c2 : 0.7x103 kg (d) Tie down device d : 1.9x103 kg (e) Lifting device e : 0.2x103 kg element 2) is called the follower type fuel element)

The basket for MNU type fuel 160 pieces

1) JRR-3 MNU* type fuel A.11. Planned years of use

1) Planned years of use: 70 years

2) Number of times used for transport per year: once or less

3) Number of days required per transport: 365 days or less Omission C. Package description - packaging Omission C.2.3 Basket (Omission)

As shown in (I)-Fig.D.1 to (I)-Fig.D.2, the size in the section of the follower fuel elements which are loaded with the basket for box type fuel is smaller than that of the standard type fuel elements. A large space (19mm when the fuel element is put aside in the lodgement.) arises between the inside of the basket lodgment and the fuel elements.Therefore, when follower type fuel elements are loaded in a basket for box type fuel, an aluminum alloy spacer shown in (I)-Fig.

C.23 shall be placed between a fuel element and a fuel element insertion hole for the purpose of using it as a heat transfer component and reducing fuel shaking during transportation.

Omission C.3.4 Basket (for box type fuel)

(1) Neutron poison Boron carbide (2) Frame Stainless steel (3) Bottom plate Stainless steel (4) Compartment plate Stainless steel, Boral plate (5) Partition plate Stainless steel (6) Adapter Aluminum alloy Omission C.5. Weight (a) Body (with fins) a : 13.8x103 kg (b) Lid, lid bolt (1) Lid (with fins) b1 : 6.7x103 kg (2) Lid bolt b2 : 0.2x103 kg (c) Basket (1) Basket for box type fuel c1 : 2.1x103 kg (2) Basket for MNU type fuel c2 : 0.7x103 kg (3) Spacer (40 pieces) c3 : 0.13x103 kg (d) Tie down device d : 1.9x103 kg (e) Lifting device e : 0.2x103 kg drawings Addition of evaluation of Consideration of aging of nuclear fuel package due to the revision of the regulations Modification for proper description Modification for proper description Addition of spacer weight

Comparison Table of SAR for Type JRC-80Y-20T before after note (f) Fuel element (1) Standard type fuel element 40 pieces f1 : 0.4x103 kg (2) Follower type fuel element (with the adapters) 40 pieces f2 : 0.4x103 kg (3) JRR-3 MNU type fuel element 160 pieces f3 : 1.6x103 kg Omission D. Contents of package Contents of the package are spent fuel elements of JRR-3 and JRR-4.

The fuel meats of uranium aluminum or uranium silicone aluminum dispersion type or uranium aluminum alloy in the fuel plates of standard type and follower type fuel element are all covered with aluminum alloy.

The metallic natural uranium of JRR-3 MNU type fuel element is covered with aluminum alloy. The fuel elements are shown in ()-Fig.D.1 through () -Fig.D.8.

The standard type fuel element, the follower type fuel element and JRR-3 MNU type fuel element are shown in ()-Fig.D.1 through () -Fig.D.4, ()-Fig.D.5 through () -

Fig.D.6 and ()-Fig.D.7 through () -Fig.D.8, respectively.

The standard type fuel elements except JRR-4 high enrichment instrumented fuel (HEU) are cut off its top and bottom portions which do not contain uranium, before being loaded in the packaging. JRR-4 high enrichment instrumented fuel (HEU) is cut off only its bottom portion which does not contain uranium, before being loaded in the packaging. The follower type fuel elements are not cut off, and are loaded in the basket.

JRR-3 MNU type fuels elements are cut into 3 pieces from the connection, and are loaded in the basket.

()-Fig.D.1 JRR-3 Standard Aluminide Type Fuel

()-Fig.D.2 JRR-3 Standard Silicide Type Fuel

()-Fig.D.3 JRR-4 Low Enrichment Silicide Type Fuel

()-Fig.D.4 JRR-4 High Enrichment Instrumented Fuel (HEU)

()-Fig.D.5 JRR-3 Follower Aluminide Type Fuel

()-Fig.D.6 JRR-3 Follower Silicide Type Fuel

()-Fig.D.7 JRR-3 MNU Type Fuel (Top, Middle Fuel)

()-Fig.D.8 JRR-3 MNU Type Fuel (Bottom Fuel)

(Fig. omitted)

(f) Fuel element (1) Standard type fuel element 40 pieces f1 : 0.4x103 kg (2) Follower type fuel element (with the adapters) 40 pieces f2 : 0.4x103 kg (3) JRR-3 MNU type fuel element 160 pieces f3 : 1.6x103 kg Omission D. Contents of package Contents of the package are spent fuel elements of JRR-3.

The fuel plates of standard type and follower type fuel elements are the fuel meats of uranium silicone aluminum dispersion type alloy covered with aluminum alloy.

The metallic natural uranium of JRR-3 MNU type fuel element is covered with aluminum alloy. The fuel elements are shown in ()-Fig.D.1 through () -Fig.D.4. The standard type fuel element, the follower type fuel element and JRR-3 MNU type fuel element are shown in

()-Fig.D.1, ()-Fig.D.2 and ()-Fig.D.3 through () -Fig.D.4 respectively.

The standard type fuel elements are cut off its top and bottom portions which do not contain uranium, so as to be a prescribed length, before being loaded in the container, and then inserted into the fuel basket.

The follower type fuel elements are not cut off, and are loaded in the basket. JRR-3 MNU type fuels elements are cut into 3 pieces from the connection, and are loaded in the basket.

()-Fig.D.1 JRR-3 Standard Silicide Type Fuel

()-Fig.D.2 JRR-3 Follower Silicide Type Fuel

()-Fig.D.3 JRR-3 MNU Type Fuel (Top, Middle Fuel)

()-Fig.D.4 JRR-3 MNU Type Fuel (Bottom Fuel)

(Fig. omitted)

Deletion of JRR-4 Modification for proper description due to deletion of fuel Changes of drawing number due to deletion of drawings Deletion of JRR-4 Deletion of JRR-3 Aluminide Changes of drawing number due to deletion of drawings Deletion of JRR-4 Changes of drawing number due to deletion of drawings

Compari son Tabl e of SAR f Or T) pe J RC-80Y20T

¨

g

¨

bef or(,

: REt t i l : ( l a, d di l i ci de t ynde1 8 011erd cr 11l owe, [ l i ci d t yp l uel el emert Han

(, add i oECt i l eJ i n one t !

) o! l

1) Thg v i t t i n t hJ l uci r`sp! c' l i at i o how u, per v`whi t h co! t d i l 1! i cn i 1l t l o( 11t t

2) Bun u

(

( ( Al l d

, 1t i ch wi l hi 1 j )

( I ni t i J l wci xh J nt ) ) x 190

3) Tl di, oel, si On of t h c( ) 1 t J 10Cd l ucl i s wi hi n t rl di mc! i pel i cd i n( 1) - 1' i l D l t hi o=h( 1), i t i : Dl

( I ) Tabl e I ) l Speci acat i on oFcont ent s

( kE/pi ccc)

( ! )

( /p`l ckage)

( P q/

ekagc)

( d`l y)

nt t P

( g/pi ccc) l

( /pi ece) l

( %) 1)

( pi cce)

J RR-3 RR-3

, RR-3 af t er Modi acat i On f or proper descri pt i on due t o del et i on of RR 3 al unni de f uel and del et i on of J RR 4 f uel

Compari son Tabl e of SAR f Or t t pe RC-8020T

¨

bef or(,

( I ) Tabl e D 2 0uant i t i es of mai o4 radi onuchdes t t e4 packaFre)

Tot al el ement Nobl e gas, et c

- 3 Kr-85 1 - 129 1-131 Xe-131m Heavy el ement Pu-238 Pu-239 Pu-240 Pu-241 Am-241 Cm-242 Cm-244 RP Sr-89 Sr-90 Y-90 Y-91 Zr-95 Nb-95 Ru-103 Ru-106 Te-129m

( Ds-134 Cs-137 Ba-140 Ce-141 Pr-143

( De-144 Pm-147 Ot hers Cl aa cat i on af t er React or Basket St andard si hci de t pe RR 3

Box t, c Fouower dt t ci de t ) e RR 3

Box t, e( t h Adapt ers)

MNUt, c RR 3 MNUt pe Modcat i On f or proper descri pt on due t o del et i on of J RR 3 al umi ni de f uel and del et i on of RR 4 f uel

Comparison Table of SAR for Type JRC-80Y-20T Before after note Chapter : Safety analyses This package is designed to comply with the IAEA Regulations for the Safe Transport of Radioactive Material 2012 Edition concerning Type B(U) package containing fissile material. This chapter shows the summary of each analysis for the package.

(1) Structural analysis In the structural analysis of the package, the evaluation of the thermal stress and internal pressure under normal and accident conditions of transport is performed by means of finite element method code ABAQUS etc. Also, the evaluation for drop tests is performed by numerical analysis using finite element method code LS-DYNA.

As the results, it is confirmed that the package sufficiently satisfies all the requirements specified in the regulation. Namely, the following is assured.

Even if the package is subject to the pressure difference between inside and outside of the packaging and thermal loads under normal and accident conditions of transport, the package has sufficiently the containment and shielding performance required in the regulation, and there is no change of the configuration which affects the criticality and thermal analysis.

It is confirmed that there is no deformation which affects criticality analysis and thermal analysis except the basket for box type fuels which suffers a slight plastic deformation partially in 9m horizontal drop. For the basket for box type fuels which suffers a slight plastic deformation, it is confirmed that there is no influence which causes criticality in the critically analysis.

Even under test conditions for package containing fissile material, as the analytical results, there is no change of the configuration which affects the evaluation of the subcriticality except the basket for box type fuels. For the basket for box type fuels which suffers a slight plastic deformation partially under the condition, it is confirmed that there is no influence on criticality also in a criticality analysis considered the plastic deformation.

(2) Thermal analysis In the thermal analysis of the package, the temperature evaluation of each portion under normal and accident conditions of transport is performed by using ABAQUS code.

As the results, it is confirmed that the package satisfies the criteria specified in the regulation under normal and accident conditions of transport, and the temperature of each portion does not affect significantly the structural strength, the containment and shielding performance.

(3) Containment analysis In the containment analysis of the package, the radioactive concentration in air inside the package is assumed to be 3.7x10-6TBq/m3 and the leakage value of radioactive gas is obtained by using the equation shown in ANSI - N 14.5 1997 edition.

As the results, it is confirmed that the obtained leakage values are very smaller than the criteria of radioactive material leakage value specified in the regulation.

(4) Shielding analysis In the shielding analysis of the package, the source intensity of the package is Chapter : Safety analyses This package is designed to comply with the IAEA Regulations for the Safe Transport of Radioactive Material 2012 Edition concerning Type B(U) package containing fissile material. This chapter shows the summary of each analysis for the package.

(1) Structural analysis In the structural analysis of the package, the evaluation of the thermal stress and internal pressure under normal and accident conditions of transport is performed by means of finite element method code ABAQUS etc. Also, the evaluation for drop tests is performed by numerical analysis using finite element method code LS-DYNA.

As the results, it is confirmed that the package sufficiently satisfies all the requirements specified in the regulation. Namely, the following is assured.

Even if the package is subject to the pressure difference between inside and outside of the packaging and thermal loads under normal and accident conditions of transport, the package has sufficiently the containment and shielding performance required in the regulation, and there is no change of the configuration which affects the criticality and thermal analysis.

It was confirmed that there is no deformation affecting the configuration which is the base for criticality analysis and thermal analysis, except the basket for box type fuel which suffers a slight plastic deformation partially in 9 m horizontal drop of the package.

It was confirmed that even under test conditions for package containing fissile material, as the analytical results, there is no impact on the configuration which is the base for the evaluation of the subcriticality, except the basket for box type fuel.

(2) Thermal analysis In the thermal analysis of the package, the temperature evaluation of each portion under normal and accident conditions of transport is performed by using ABAQUS code.

As the results, it was confirmed that the package satisfies the criteria specified in the regulation under normal and accident conditions of transport, and the temperature of each portion does not significantly affect the structural strength, the containment and shielding performance.

(3) Containment analysis In the containment analysis of the package, the radioactive concentration in air inside the package is assumed to be 3.7x10-6TBq/m3 and the leakage value of radioactive gas is obtained by using the equation shown in ANSI - N 14.5 1997 edition.

As the results, it was confirmed that under normal and accident conditions of transport, the obtained leakage values satisfied the criteria of radioactive material leakage value specified in the regulation and notification.

(4) Shielding analysis In the shielding analysis of the package, the source intensity of the package is Modification for proper description Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before after note calculated by using the ORNL isotope generation and depletion code ORIGEN or ORIGEN-JR. Also, the gamma and neutron shielding calculation are performed by using the point kernel code QAD-CGGP2R and two dimensional discrete ordinates transport code DOT 3.5, respectively.

As the results, it is confirmed that the dose equivalent ratios on the surface of the package and at points of 1 m from its surface are very small as compared with the criteria of each case specified in the regulations under routine, normal and accident conditions of transport.

(5)

Criticality analysis The criticality analysis is performed by using the three dimensional multigroup Monte Carlo KENO-Va code.

As the results, it is confirmed that the subcriticality of the package in isolation is maintained under routine, normal and accident conditions of transport pertaining to package containing fissile material. And it is also confirmed that the subcriticality of the package arrays is maintained under normal and accident conditions of transport pertaining to package containing fissile material.

calculated by using the ORNL isotope generation and depletion code ORIGEN or ORIGEN-JR. Also, the gamma and neutron shielding calculation are performed by using the point kernel code QAD-CGGP2R and two dimensional discrete ordinates transport code DOT 3.5, respectively.

As the results, it was confirmed that the dose equivalent ratios on the surface of the package and at points of 1 m from its surface are sufficiently small compared with the criteria of each case specified in the regulations under routine, normal and accident conditions of transport.

(5)

Criticality analysis The criticality analysis is performed by using the three dimensional multigroup Monte Carlo KENO-Va code.

As the results, it was demonstrated that there is no deformation, etc. of the structure affecting subcriticality evaluation under normal conditions of transport pertaining to package containing fissile material, and it was confirmed that the subcriticality is maintained for the nuclear fuel package under routine conditions of transport, for the nuclear fuel package in isolation, and for the package in isolation and in array under the normal and accident conditions of transport pertaining to package containing fissile material.

(6)

Consideration of Aging of Nuclear Fuel Package As a result of the evaluation of aging effects due to the factors such as heat, radiation, and chemical changes under the conditions of use expected during the planned period of use, it was confirmed that such effects need not be considered in confirming compliance with the technical criteria. For the lifting and containment devices, it is necessary to consider aging effects due to fatigue as stress will be generated repeatedly. For the lifting and containment devices, each fatigue was evaluated considering the conservative repeat count expected during the period of use, and it was confirmed that there is no impact on compliance with the technical criteria as fatigue failure does not occur.

The details of each analysis and the evaluations are described in Chapter A through Chapter F.

For the purpose of conservative evaluation, the safety analysis assumes the cases where following fuel elements are loaded, which will pose more severe conditions than the current contents.

Same as above Same as above Addition of evaluation of Consideration of aging of nuclear fuel package due to the revision of the regulations Modification for proper description Changes of description due to deletion of JRR-3 aluminide fuel and

Comparison Table of SAR for Type JRC-80Y-20T Before after note Fuel Element A: JRR-3 Standard Aluminide Type Fuel Element (decay heat, maximum temperature)

Fuel Element B: JRR-4 Low Enrichment Type Fuel Element (LEU) (Neutron source intensity)

JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note A. Structural analysis omission A. Structural analysis omission Deletion due to re-evaluation

Comparison Table of SAR for Type JRC-80Y-20T Before After note Deletion of JRR-3 aluminide fuel and change due to moving from separate items descriptions

Comparison Table of SAR for Type JRC-80Y-20T Before After note Move of descriptions to separate lines and deletion of JRR-3 aluminide fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note Deletion of JRR-4 fuel Changes of drawing number

Comparison Table of SAR for Type JRC-80Y-20T Before After note Deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note Deletion of JRR-3 aluminide fuel Changes of drawing number

Comparison Table of SAR for Type JRC-80Y-20T Before After note Deletion of JRR-3 aluminide fuel and JRR-4 fuel Deletion due to moving to separate items descriptions Changes of drawing number

Comparison Table of SAR for Type JRC-80Y-20T Before After note Changes of drawing number

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission omission Modification for proper description Deletion of JRR-3 aluminide fuel and JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission A.4.2 Low-temperature strength omission a

Clearance between the shell and the basket under the ambient temperature -40.

The temperature distribution obtained by thermal analysis under ambient temperature -40 of the package is applied. The temperature distribution of the inner and the outer surface of the shell is shown in (II)-Fig A.12, and the temperature distribution of the basket is shown in (II)-Fig A.13. (II)-Fig.A.13 shows the temperature distribution of the basket for box type fuel, which has the maximum expansion value, in the case of JRR-3 standard aluminide type fuel being contained.

(omission

) Clearance in the radial direction r between the shell and the basket Accordingly, the clearance r between the body and the basket is given as follows ;

r = 10.2640.492 = 0.244 (mm)

Therefore, a clearance between the body and the basket at the ambient temperature of -40 is 0.244 mm.

No stress due to constraint will occur in each basket.

omission A.4.4 Lifting device omission In the design fatigue curve, the allowable number of cycles N corresponding to the value S'alt is as follows.

N = 7.4x105 times On the other hand, the number of cycles n of the lifting lug used during the life of the packaging is as follows.

n =x104 times Therefore, the safety factor (RF) and the safety margin (MS) of the fatigue strength of the body lifting lug are as follows.

RF =

4 5

10 x

1 10 x

4 7

= 74 MS = 741 = 73 Therefore, The integrity of the body lifting lug can be assured, since the allowable number of cycles of the body lifting lug N is sufficiently greater than the number of cycles during the life of the packaging n.

omission A.4.2 Low-temperature strength omission a

Clearance between the shell and the basket under the ambient temperature -40.

The temperature distribution obtained by thermal analysis under ambient temperature of -40 of the nuclear fuel package is applied. The temperature distribution of the inner and the outer surface of the shell is shown in (II)-Fig A.12, and the temperature distribution of the basket is shown in (II)-Fig A.13. Here, for the basket, in order to conduct a more conservative evaluation than in the case of the contents, it shows the temperature distribution of the basket for box type fuel containing a fuel element assumed to have the largest expansion amount under normal conditions (hereinafter referred to as fuel element A). Therefore, the evaluation for the relation between the displacement (expansion) of the basket and the displacement (shrink) of the shell is made as follows.

omission

) Clearance in the radial direction r between the shell and the basket Accordingly, the clearance r between the body and the basket is given as follows ;

r = 10.2640.492 = 0.244 (mm)

Therefore, a clearance between the body and the basket at the ambient temperature of -40 is 0.244 mm. And even when considering the expected ambient temperature change during transportation (from -40 to 38), since there will be no change in the temperature range for the same material, there is no difference in the expansion amount of the transport container body nor that of the fuel basket, and no change in the amount of gap either.

No stress due to constraint will occur in each basket.

omission A.4.4 Lifting device omission In the design fatigue curve, the allowable number of cycles N corresponding to the value S'alt is as follows.

N = 7.4x105 times On the other hand, assuming that the expected service life is 70 years, the frequency of use is once a year, and the number of handling times per transportation is 100 times, the repeat count of lifting (n) will be n = 7000 times. Here, to conservatively consider the repeat count of lifting, the value to be used in the calculation will be n =x104 times Therefore, the safety factor (RF) and the safety margin (MS) of the fatigue strength of the body lifting lug are as follows.

RF =

4 5

10 x

1 10 x

4 7

= 74 MS = 741 = 73 Modification for proper description due to deletion of JRR-3 aluminide fuel Addition of evaluation of Consideration of aging of nuclear fuel package due to revision of the regulations Modification for proper description on the evaluation of "Consideration of aging of nuclear fuel package" due to revision of the regulations

Comparison Table of SAR for Type JRC-80Y-20T Before After note Therefore, The integrity of the body lifting lug can be assured, since the allowable number of cycles of the body lifting lug N is sufficiently greater than the number of cycles during the life of the packaging n. Based on the above, as a result of the fatigue evaluation by setting the repeat count conservatively, it was confirmed that fatigue failure did not occur.

Modification for proper description

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission A.4.6 Pressure omission The change on the stress corresponding to the increase in the internal pressure is obtained from the result of stress calculation due to only the internal pressure in paragraph A.5.1.3. The maximum stress (stress intensity) occurs in the inner surface at the center of the body bottom plate, and the value is 0.077 MPa (= 0.06x2.270.06).

Therefore, if the ambient pressure is reduced to 6.0x104 Pa, the effect of the pressure reduction upon the packaging can be ignored.

As shown in paragraph ()-C.3.1 2, the leaktightness test of the packaging is performed at the internal pressure of greater than 0.42 MPaG and has been verified that the containment of the packaging is secured. Therefore, the safety of the packaging can be secured without affecting its containment.

A.4.7 Vibration omission Therefore, 1 999 Hz Since the natural frequency of the package is about 999 Hz against the maximum effective frequency (approx. 50 Hz)15) which is predicted during transport, there will be no resonance during transport.

Therefore, various bolts used for the packaging will not cause resonance, and they will not become loose because they are provided with antirotation keys.

omission A.4.6 Pressure omission The change on the stress corresponding to the increase in the internal pressure is obtained from the result of stress calculation due to only the internal pressure in paragraph A.5.1.3. The maximum stress (stress intensity) occurs in the inner surface at the center of the body bottom plate, and the value is 0.077 MPa (= 0.06x2.270.06).

Therefore, if the ambient pressure is reduced to 6.0x104 Pa, the effect of the pressure reduction upon the packaging can be ignored.

In addition, in the evaluation of the strain level of the lid sealing boundary for the special test conditions shown in (II)-Table A.21, it was confirmed that the sealing property was maintained since the recovery of the initial clamping stress was confirmed and the mouth opening was less than 2.01 mm for an initial clamping allowance of 3 mm.

A.4.7 Vibration omission Therefore, f1 = 999 Hz Since the natural frequency of the package is about 999 Hz against the maximum effective frequency (approx. 50 Hz)15) which is predicted during transportation, there will be no resonance during transportation.

The relationship24) between the amplification factor and the ratio of frequencies is given by the curve shown in (II)-Fig. A.34. Here, since the expected frequency of vibrations during transportation is about 50 Hz, the amplification factor will be obtained as follows:

Ratio of frequencies f / fn 0.05 Where, fnNatural frequency of the package f: Frequency of vibrations during transportation Therefore, Amplification factor 1 Therefore, since there is no influence of load amplification due to vibrations during transportation, in addition, in the stacking evaluation ((II)-A.5.4) under the normal test conditions, considering the fact that the transport container will not be deformed even when it is subjected to five times its own weight + its own weight load, there is no risk of cracks, damages, etc. to the transport container due to vibrations during transportation.

Modification for proper description due to re-evaluation Modification for proper description Modification for proper description on the evaluation of "Consideration of aging of nuclear fuel package" due to revision of the regulations

Comparison Table of SAR for Type JRC-80Y-20T Before After note

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission

.5.1 Thermal test The results of thermal analysis of the package conducted under the normal conditions of transport are summarized as shown in (II)-B.4.

1)

The maximum temperature and the maximum internal pressure are generated in an environment under the solar insolation when JRR-3 standard aluminide type fuel with the maximum decaying heat is contained.

omission

.5.1.1 Summary of pressure and temperature omission The temperature distribution of the packaging and basket is shown in (II)-Fig.A.34 through Fig.A.36. The temperature distribution of the components of the packaging shown in (II)-Fig.A.34 denotes the values of the maximum temperature gradient when fuel elements are contained in the packaging.

The contained fuel is the JRR-3 standard aluminide type fuel which is thermally analyzed under the conditions in an environment at 38 without the solar insolation.

omission

.5.1.2 Thermal expansion

1.

Thermal stress and deformation of the packaging omission The maximum internal pressure 5.17'104 PaG (the value of the basket for box type fuel, when the JRR-3 standard aluminide type fuel is contained, shown in (II)-Table A.8),

is rounded conservatively as 6.0 '104 PaG. Temperature distribution is used in which the greatest thermal stress is expected. Thus, the value having the greatest temperature gradient (when under 38°C of ambient temperature and JRR-3 standard aluminide type fuel is contained in the packaging the absence of insolation) is used.

omission The result of the analysis is as follows;.

b <0 + body (1.44 mm < 4.526 mm)

Therefore, it becomes the following in any basket.

b <0 + body omission

.5.1 Thermal test The results of thermal analysis of the nuclear fuel package conducted under the normal test conditions are summarized as shown in (II)-B.4.

1) The maximum temperature and maximum internal pressure will occur, when considering conservatively, under the environment where the fuel element As which have the greatest decay heat are contained and are subjected to the solar radiation heat.

omission

.5.1.1 Summary of pressure and temperature omission The temperature distribution of the packaging and basket is shown in (II)-Fig.A.34 through Fig.A.36. The temperature distribution of the components of the packaging shown in (II)-Fig.A.34 denotes the values of the maximum temperature gradient when fuel elements are contained in the packaging.

The contained fuel was the fuel element A which was thermally analyzed under the conditions in an environment at 38 without the solar insolation.

omission

.5.1.2 Thermal expansion

1.

Thermal stress and deformation of the packaging omission The maximum internal pressure of 5.17x104 PaG when the fuel element A is contained (the value of the basket for box type fuel shown in (II)-Table A.8) was rounded conservatively as 6.0 x104 PaG. The temperature distribution in which the greatest thermal stress would be expected, that is, temperature gradient would be the greatest (under ambient temperature of 38°C, without solar insolation and containing fuel element A) was used.

omission The result of the analysis is as follows;.

b <0 + body (1.44 mm < 4.526 mm)

Modification for proper description due to deletion of JRR-3 aluminide fuel same as above same as above same as above same as above

Comparison Table of SAR for Type JRC-80Y-20T Before After note Therefore, the basket is not restricted between the bottom plate of the body and the internal surface of the lid. Therefore, the stress generated in the basket is caused by the temperature gradient of the basket itself.

Therefore, it becomes the following in any basket.

b <0 + body And even when considering the expected ambient temperature changes during transportation (from -40 to 38), since there is no change in the temperature range of the same material, there is no difference in the expansion amount of the transport container body nor that of the fuel basket, and no change in the amount of gap either.

Therefore, the basket is not restricted between the bottom plate of the body and the internal surface of the lid. Therefore, the stress generated in the basket is caused by the temperature gradient of the basket itself.

Addition of evaluation of Consideration of aging of nuclear fuel package due to revision of the regulations

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission As a result of the calculation obtained by the above expressions, the elongations rb in the radial direction of the basket are shown in (II)-Table A.12. In the Table, the initial gap g0 at the room temperature, the safety factor, and the safety margin are also shown.

Therefore, as apparent from the Table, there is still a gap between each basket and the internal surface of the packaging even after thermal expansion. For this reason, the stress due to restriction is not generated in each basket.

omission A.5.1.4 Comparison of allowable stress (omission Therefore, the safety factor (RF) and the safety margin (MS) on the stress classification line A of the body and the lid are as follows; RF = 411 59.4 = 6.9 MS = 6.9 1 = 5.9 Therefore, the structural integrity of the packaging is maintained, because the stress which is generated on each part of the packaging is within the allowable stress.

omission On the other hand, the number of operating cycles n during the life of the packaging is 100 cycles.

The safety factor (RF) and the safety margin (MS) of the fatigue strength of the lid bolt are as follows; RF = 3400 100 = 34 MS = 34 1 = 33 Therefore, the allowable number of cycles of the lid bolt is sufficiently greater than the number of operating cycles during the life of the packaging, and the fatigue failure dose not occur in the lid bolt.

omission As a result of the calculation obtained by the above expressions, the elongations rb in the radial direction of the basket are shown in (II)-Table A.12. In the Table, the initial gap g0 at the room temperature, the safety factor, and the safety margin are also shown.

Therefore, as apparent from the Table, there is still a gap between each basket and the internal surface of the packaging even after thermal expansion. For this reason, the stress due to restriction is not generated in each basket.

And even when considering the expected ambient temperature changes during transportation (from -40 to 38), since there is no change in the temperature range of the same material, there is no difference in the expansion amount of the transport container body nor that of the fuel basket, and no change in the amount of gap either.

omission A.5.1.4 Comparison of allowable stress omission Therefore, the safety factor (RF) and the safety margin (MS) on the stress classification line A of the body and the lid are as follows;

= 411 59.4 = 6.9

= 6.9 - 1 = 5.9 Therefore, the structural integrity of the packaging is maintained, because the stress which is generated on each part of the packaging is within the allowable stress.

In addition, the temperature change from room temperature (20) to 105 was studied. When considering the lowest ambient temperature of -40 expected during transportation, the temperature difference between -40 and 105 is 145. The stress in this case will be 59.4 x 145/85 101.3 102 MPa Therefore, the safety factor RF and the margin factor MS are as follows:

= 411 102 = 4.0

= 4.0 - 1 = 3.0 Then, there is no risk of cracks or failures, etc. even when ambient temperature changes expected during transportation are considered.

omission On the other hand, the number of operating cycles n during the life of the packaging is 300 cycles.

The safety factor (RF) and the safety margin (MS) of the fatigue strength of the lid bolt are as follows;

= 3400 300 = 11.3

= 11.3 - 1 = 10.3 Addition of evaluation of Consideration of aging of nuclear fuel package due to revision of the regulations Addition of evaluation of Consideration of aging of nuclear fuel package due to revision of the regulations Changes due to re-evaluation of "Consideration of aging of nuclear fuel package" due to revision of the regulations

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission A.6.1 Mechanical test drop test I (9 m drop) omission (2) The total mass of the package is a maximum when the basket for box type fuel such as JRR-3 standard aluminide type fuel is contained. The maximum total mass of the package 23.2 x 103 kg is used as the total mass in the analysis.

omission A.6.1.1 Vertical drop omission

1. Containment at the contact surface of the lid and the body 1.1 Top vertical drop 1.2 Bottom vertical drop

2. Strength of the valve

3. Strength of the baskets 3.1 Basket for box type fuel 3.2 Basket for MNU type fuel

4. Strength of the fuel elements 4.1 JRR-3 standard aluminide type fuel 4.2 JRR-3 standard silicide type fuel 4.3 JRR-3 follower aluminide type fuel 4.4 JRR-3 follower silicide type fuel 4.5 JRR-3 MNU type fuel 4.6 JRR-4 low enrichment silicide type fuel 4.7 JRR-4 high enrichment instrumented fuel (HEU) omission 3.1 Strength of the basket for box type fuel 3.1.1 Strength of the basket omission Accordingly, the safety factor (RF) and the safety margin (MS) against buckling are as follows; RF = 801 40.0 = 20 M.S = 201 = 19 Therefore, no buckling will occur in the basket for box type fuel due to the vertical drop.

Therefore, the allowable number of cycles of the lid bolt is sufficiently greater than the number of operating cycles during the life of the packaging, and the fatigue failure dose not occur in the lid bolt.

Times of use N = 4/yearx70 yearsxmargin 300 times omission A.6.1 Mechanical test drop test I (9 m drop) omission (2) The total mass of the package will be a maximum when the basket for box type fuel (contents: JRR-3 standard Silicide type fuel element, etc.) is contained. The maximum total mass of 23.2 x 103 kg is used in the analysis.

omission A.6.1.1 Vertical drop omission

1. Containment at the contact surface of the lid and the body 1.1 Top vertical drop 1.2 Bottom vertical drop

2. Strength of the valve

3. Strength of the baskets 3.1 Basket for box type fuel 3.2 Basket for MNU type fuel

4. Strength of the fuel elements 4.1 JRR-3 standard silicide type fuel 4.2 JRR-3 follower silicide type fuel 4.3 JRR-3 MNU type fuel omission 3.1 Strength of the basket for box type fuel 3.1.1 Strength of the basket omission Accordingly, the safety factor (RF) and the safety margin (MS) against buckling are as follows;

= 801 40.0 = 20

= 20 - 1 = 19 Therefore, no buckling will occur in the basket for box type fuel due to the vertical drop. In addition, as shown in "(I) C. Transport Container," the neutron absorber is surrounded by basket dividers, which means it will not be crushed and have no effect on subcriticality.

Modification for proper description due to deletion of JRR-3 aluminide fuel Change of item number due to deletion of JRR-3 aluminide fuel same as above Deletion of JRR-4 fuel Modification for proper description

Comparison Table of SAR for Type JRC-80Y-20T Before After note 3.1.2 Strength of the neutron poison The neutron poison is as shown in (II)-Fig.A.81.

It has length of 925mm, width of 314.5mm and thickness of 4.5mm.

In the case of vertical drop, the compressive stress occurs in the neutron poison due to inertia force.

This compressive stress (c) is given by the following formula.

c = F A = WgGv A

(II)-Fig.A.81 Analytical model of the neutron poison of the basket for box type fuel where, c

Compressive stress (MPa)

A

Pressure area of the neutron poison to the vertical direction

= 314.5 x 4.5 =1.41 x 103 (mm2)

F

Inertia force of the neutron poison

=WgGv (N)

W

Mass of the neutron poison Since the density of neutron poison () is 2.67 x 10-6 kg/ mm3, the mass is given as follows; W = 2.67 x 10-6 x 925 x 314.5 x 4.5 = 3.50 (kg) g

Gravitational acceleration

= 9.8 (m/sec2)

Gv

Impact deceleration = 445 (g)

Therefore, the compressive stress (c) on the neutron poison is given as follows; c = 3.50 x 9.8 x 445 1.41 x 103

= 10.9 MPa The allowable compressive stress (ca) of the neutron poison is given as follows; ca= 1.5 x y = 1.5 x 44.1 = 66.1 MPa Therefore, the safety factor (RF) and the safety margin (MS) of neutron poison are as follows; RF = 66.1 10.9 = 6.0 MS = 6.01 = 5.0 Consequently, the neutron poison is never crushed due to the vertical drop, and no influence is brought about to the subcriticality.

Deletion due to re-evaluation 4.5 F

925 (Unit : mm)

F 314.5

Comparison Table of SAR for Type JRC-80Y-20T Before After note

4. Strength of the fuel element 4.1 Strength of JRR-3 standard aluminide type fuel The inertia force due to the vertical drop acts on the fuel element at the drop.

This paragraph shows that the fuel side plate has the sufficient strength enough to be resistible against the such inertia force, and further that the fuel plate never fall.

4.1.1 Strength of the fuel side plate The compressive stress (c) generated at the fuel side plate due to drop is given by the following equation.

c = F A = WgGv A

where, c

Compressive stress (MPa)

A

Pressure area of the fuel side plate to the vertical direction

= 76.2 x 4.8 x 2 = 7.31 x 102 (mm2)

F

Inertia force of the fuel element = WgGv (N)

W

Mass of the fuel element = 8 (kg) g

Gravitational acceleration = 9.8 (m/sec2)

Gv

Impact deceleration = 445 (g)

Consequently, the compressive stress (c) on the fuel side plate is as follows; c = 8 x 9.8 x 445 7.31 x 102 = 47.8 MPa In this case, the yield stress (y) of 46 MPa of material A6061-T6 is used, since the yield stress of material A6061-T6 at the operating temperature of 240 is less than that of material AG3NE as shown in (II)-Fig.A.6.

Therefore, the allowable compressive stress (ac) of fuel element of the temperature of 240 is given as follows; ac = 1.5 x y = 1.5 x 46 = 69 MPa Therefore, the safety factor (RF) and the safety margin (MS) are as follows; RF = 69 47.8 = 1.4 MS = 1.41 = 0.4 Therefore, the fuel side plate has the sufficient strength against the compressive stress due to the vertical drop.

4.1.2 Evaluation of the fuel plate for falling The fuel plate is fixed to two fuel side plates by using ro11 swage method.

The impact force (F) applied to the fuel plate due to the vertical drop is given by the following equation; F = W x g x Gv

where, W

Mass of fuel element = 8.0 (kg)

Gv

Maximum deceleration at the time of vertical drop = 445 (g)

Deletion of JRR-3 aluminide fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note Consequently, the following is given; F = 8.0 x 9.8 x 445 = 3.49 x 104 N On the other hand, the anchoring force (Fr) of fuel plate fixed to the fuel side plate by using roll swage method is given as follows; Fr = nLq

where, n

Number of fuel plate = 20 (pieces)

L

Length of the fuel plate = 770 (mm) q

Anchoring force of the fuel plate for each unit length in the fuel production specification = more than 26.5 (N/mm)

Consequently, the following is given; Fr = 20 x 770 x 26.5 = 4.08 x 105 N Therefore, the safety factor (RF) and the safety margin (MS) against fixation of the fuel plate are as shown below; RF = Fr F =

4.08 x 105 3.49 x 104 = 11 MS = 111 = 10 Therefore, the fuel plate is retained without falling.

4.2 Strength of JRR-3 standard silicide type fuel The inertia force due to the vertical drop acts on the fuel element in the same manner as the JRR-3 standard aluminide type fuel.

This paragraph shows that the fuel side plate has the sufficient strength enough to be resistible against the such inertia force, and further that the fuel plate never fall.

4.2.1 Strength of the fuel side plate No change 4.2.2 Evaluation of the fuel plate for falling (No change 4.3 Strength of JRR-3 follower aluminide type fuel The inertia force acts on the fuel element in the case of the vertical drop in the same manner as JRR-3 standard aluminide type fuel.

This paragraph shows that the fuel side plate has the sufficient strength enough to be resistible against the such inertia force, and further that the fuel plate never fall.

4.3.1 Strength of the fuel side plate The compressive stress (c) generated at the fuel side plate due to drop is given by the following equation.

c = F A = WgGv A

where, c

Compressive stress (MPa) 4.1 Strength of JRR-3 standard silicide type fuel The inertia force due to the vertical drop acts on the fuel element.

This paragraph shows that the fuel side plate has the sufficient strength enough to be resistible against the such inertia force, and further that the fuel plate never fall.

4.1.1 Strength of the fuel side plate (No change 4.1.2 Evaluation of the fuel plate for falling No change Changes of item number and description due to deletion of JRR-3 aluminide fuel same as above same as above Deletion of JRR-3 aluminide fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note A

Pressure area of the fuel side plate to the vertical direction

= 63.6 x 4.8 x 2 = 6.10 x 102 (mm2)

F

Inertia force of the fuel element = WgGv (N)

W

Mass of the fuel element = 5.2 (kg) g

Gravitational acceleration = 9.8 (m/sec2)

Gv

Impact deceleration = 445 (g)

Consequently, the compressive stress (c) on the fuel side plate is as follows; c = 5.2 x 9.8 x 445 6.10 x 102

= 37.2 MPa In this case, the yield stress (y) of 46 MPa of material A6061-T6 is used, since the yield stress of material A6061-T6 at the temperature of 240 is smaller than that of material AG3NE as shown in (II)-Fig.A.6.

Therefore, the allowable compressive stress (ac) of fuel element of the temperature of 240 is given as follows; ac = 1.5 x y = 1.5 x 46 = 69 MPa Therefore, the safety factor (RF) and the safety margin (MS) are as follows; RF = 69 37.2 = 1.8 MS = 1.81 = 0.8 Therefore, the fuel side plate has the sufficient strength against the compressive stress due to the vertical drop.

4.3.2 Evaluation of the fuel plate for falling The fuel plate is fixed to two fuel side plates by using ro11 swage method.

The impact force (F) applied to the fuel plate due to the vertical drop is given by the following equation; F = W x g x Gv

where, W

Mass of fuel element = 5.2 (kg)

Gv

Maximum deceleration at the time of vertical drop = 445 (g)

Consequently, the following is given; F = 5.2 x 9.8 x 445 = 2.27 x 104 N On the other hand, the anchoring force (Fr) of fuel plate fixed to the fuel side plate by using roll swage method is given as follows; Fr = nLq

where, n

Number of fuel plate = 16 (pieces)

L

Length of the fuel plate = 770 (mm) q

Anchoring force of the fuel plate for each unit length in the fuel production specification = more than 26.5 (N/mm)

Consequently, the following is given;

Comparison Table of SAR for Type JRC-80Y-20T Before After note Fr = 16 x 770 x 26.5 = 3.26 x 105 N Therefore, the safety factor (RF) and the safety margin (MS) against fixation of the fuel plate are as shown below; RF = Fr F = 3.26 x 105 2.27 x 104

= 14 MS = 141 = 13 Therefore, the fuel plate can be retained without falling.

4.4 Strength of JRR-3 follower silicide type fuel The inertia force due to the vertical drop acts on the fuel element in the same manner as JRR-3 standard aluminide type fuel.

This paragraph shows that the fuel side plate has the sufficient strength enough to be resistible against the such inertia force, and further that the fuel plate never fall.

4.4.1 Strength of the fuel side plate No change 4.4.2 Evaluation of the fuel plate for falling No change 4.2 Strength of JRR-3 follower silicide type fuel The inertia force due to the vertical drop acts on the fuel element in the same manner as JRR-3 standard silicide type fuel element.

This paragraph shows that the fuel side plate has the sufficient strength enough to be resistible against the such inertia force, and further that the fuel plate never fall.

4.2.1 Strength of the fuel side plate No change 4.2.2 Evaluation of the fuel plate for falling No change Changes of item number and description due to deletion of JRR-3 aluminide fuel same as above same as above

Comparison Table of SAR for Type JRC-80Y-20T Before After note 4.5 Strength of JRR-3 MNU type fuel omission 4.6 Strength of JRR-4 low enrichment silicide type fuel The inertia force due to the vertical drop acts on the fuel element in the same manner as JRR-3 standard aluminide type fuel.

This paragraph shows that the fuel side plate has the sufficient strength enough to be resistible against the such inertia force, and further that the fuel plate never fall.

4.6.1 Strength of the fuel side plate The compressive stress (c) generated at the fuel side plate due to drop is given by the following equation.

c = F A = WgGv A

where, c

Compressive stress (MPa)

A

Pressure area of the fuel side plate to the vertical direction

= 80 x 4.8 x 2 = 7.68 x 102 (mm2)

F

Inertia force of the fuel element = WgGv (N)

W

Mass of the fuel element = 5.6 (kg) g

Gravitational acceleration = 9.8 (m/sec2)

Gv

Impact deceleration = 445 (g)

Consequently, the compressive stress (c) on the fuel side plate is as follows; c = 5.6 x 9.8 x 45 7.68 x 102

= 31.8 MPa In this case, the yield stress (y) of 46 MPa of material A6061-T6 is used, since the yield stress of material A6061-T6 at the temperature of 240 is smaller than that of material AG3NE as shown in (II)-Fig.A.6.

Therefore, the allowable compressive stress (ac) of fuel element of the temperature of 240 is given as follows; ac = 1.5 x y = 1.5 x 46 = 69 MPa Therefore, the safety factor (RF) and the safety margin (MS) are as follows; RF = 69 31.8 = 2.1 MS = 2.11 = 1.1 Therefore, the fuel side plate has the sufficient strength against the compressive stress due to the vertical drop.

4.6.2 Evaluation of the fuel plate for falling The fuel plate is fixed to two fuel side plates by using ro11 swage method.

The impact force (F) applied to the fuel plate due to the vertical drop is given by the following equation; F = W x g x Gv 4.3 Strength of JRR-3 MNU type fuel No change Modification for proper description due to deletion of JRR-3 aluminide fuel Deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note

where, W

Mass of fuel element = 5.6 (kg)

Gv

Maximum deceleration at the time of vertical drop = 445 (g)

Consequently, the following is given; F = 5.6 x 9.8 x 445 = 2.45 x 104 N On the other hand, the anchoring force (Fr) of fuel plate fixed to the fuel side plate by using roll swage method is given as follows; Fr = nLq

where, n

Number of fuel plate = 15 (pieces)

L

Length of the fuel plate = 630 (mm) q

Anchoring force of the fuel plate for each unit length in the fuel production specification = more than 26.5 (N/mm)

Consequently, the following is given; Fr = 15 x 630 x 26.5 = 2.50 x 105 N Therefore, the safety factor (RF) and the safety margin (MS) against fixation of the fuel plate are as shown below; RF = Fr F = 2.50 x 05 2.45 x 104 = 10 MS = 101 = 9 Therefore, the fuel plate is retained without falling.

4.7 Strength of JRR-4 high enrichment instrumented fuel (HEU)

The inertia force due to the vertical drop acts on the fuel element in the same manner as JRR-3 standard aluminide type fuel.

This paragraph shows that the fuel side plate has the sufficient strength enough to be resistible against the such inertia force, and further that the fuel plate never fall.

4.7.1 Strength of the fuel side plate The compressive stress (c) generated at the fuel side plate due to drop is given by the following equation.

c = F A = WgGv A

where, c

Compressive stress (MPa)

A

Pressure area of the fuel side plate to the vertical direction

= 80 x 4.8 x 2 = 7.68 x 102 (mm2)

F

Inertia force of the fuel element = WgGv (N)

W

Mass of the fuel element = 6.0 (kg) g

Gravitational acceleration = 9.8 (m/sec2)

Gv

Impact deceleration = 445 (g)

Consequently, the compressive stress (c) on the fuel side plate is as follows; c = 6.0 x 9.8 x 445 7.68 x 102

= 34.1 MPa Deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note In this case, the yield stress (y) of 46 MPa of material A6061-T6 is used, since the yield stress of material A6061-T6 at the temperature of 240 is smaller than that of material AG3NE as shown in (II)-Fig.A.6.

Therefore, the allowable compressive stress (ac) of fuel element of the temperature of 240 is given as follows; ac = 1.5 x y = 1.5 x 46 = 69 MPa Therefore, the safety factor (RF) and the safety margin (MS) are as follows; RF = 69 34.1 = 2.0 MS = 2.01 = 1.0 Therefore, the fuel side plate has the sufficient strength against the compressive stress due to the vertical drop.

4.7.2 Evaluation of the fuel plate for falling The fuel plate is fixed to two fuel side plates by using ro11 swage method.

The impact force (F) applied to the fuel plate due to the vertical drop is given by the following equation; F = W x g x Gv

where, W

Mass of fuel element = 6.0 (kg)

Gv

Maximum deceleration at the time of vertical drop = 445 (g)

Consequently, the following is given; F = 6.0 x 9.8 x 445 = 2.62 x 104 N On the other hand, the anchoring force (Fr) of fuel plate fixed to the fuel side plate by using roll swage method is given as follows; Fr = nLq

where, n

Number of fuel plate = 15 (pieces)

L : Length of the fuel plate = 630 (mm)

Removing 6 width 5mm cuttings for measurement on the fuel plate, the net length of fuel plate comes to L=630-6 x 5=600mm q

Anchoring force of the fuel plate for each unit length in the fuel production specification = more than 26.5 (N/mm)

Consequently, the following is given; Fr = 15 x 600 x 26.5 = 2.39 x 105 N Therefore, the safety factor (RF) and the safety margin (MS) against fixation of the fuel plate are as shown below; RF = Fr F = 2.39 x 105 2.62 x 104 = 9 MS = 91 = 8 Therefore, the fuel plate is retained without falling.

Comparison Table of SAR for Type JRC-80Y-20T Before After note

.6.1.2 Horizontal drop omission

4. Strength of the fuel elements 4.1 JRR-3 standard aluminide type fuel 4.2 JRR-3 standard silicide type fuel 4.3 JRR-3 follower aluminide type fuel 4.4 JRR-3 follower silicide type fuel 4.5 JRR-3 MNU type fuel 4.6 JRR-4 low enrichment silicide type fuel 4.7 JRR-4 high enrichment instrumented fuel (HEU) omission 3.1 Basket for box type fuel 3.1.1 Strength of the compartment plate and the partition plate The stress calculation for the horizontal drop of the basket for box type fuel is made for case when the basket drops to two directions of the X-direction and the Y-direction as shown in (II)-Fig.A.93.

The stress calculation for the basket is made by using the general-purpose finite element program ABAQUS.

The analysis is also made on the shearing strength at the weld zone between the compartment plate and the partition plate.

a) Strength for the drop to the X-direction In the analysis, the inertia force of the basket itself, the fuel element and the neutron poisons generated by the deceleration are taken into.

The partition plate and the compartment plate are modeled with the shell element. As the mass of the furl element in the analysis, the largest mass of standard type fuel element is used. It is assumed JRR-3 standard aluminide type fuel as a representative.

omission b) Strength for the drop to the Y-direction Strength evaluation at the Y-direction drop is carried out. In the analysis, it is modeled in the same manner as the X-direction drop. Namely, the partition plate and the compartment plate are modeled with the shell element, and the fuel element, whose cross-sectional form is a rectangle (76.2 mmx76.2 mm), is modeled with the solid element that has equivalent stiffness. JRR-3 standard aluminide type fuel that has the largest mass is used again as the fuel element in the analysis.

omission

.6.1.2 Horizontal drop omission

4. Strength of the fuel elements 4.1 JRR-3 standard silicide type fuel 4.2 JRR-3 follower silicide type fuel 4.3 JRR-3 MNU type fuel omission 3.1 Basket for box type fuel 3.1.1 Strength of the compartment plate and the partition plate The stress calculation for the horizontal drop of the basket for box type fuel is made for case when the basket drops to two directions of the X-direction and the Y-direction as shown in (II)-Fig.A.93.

The stress calculation for the basket is made by using the general-purpose finite element program ABAQUS.

The analysis is also made on the shearing strength at the weld zone between the compartment plate and the partition plate.

Strength for the drop to the X-direction In the analysis, the inertia force of the basket itself, the fuel element and the neutron poisons generated by the deceleration are taken into.

The partition plate and the compartment plate are modeled with the shell element. As the mass of the fuel element in the analysis, the largest mass of standard type fuel element is used. JRR-3 standard silicide type fuel is assumed as a representative.

omission b) Strength for the drop to the Y-direction Strength evaluation at the Y-direction drop is carried out. In the analysis, it is modeled in the same manner as the X-direction drop. Namely, the partition plate and the compartment plate are modeled with the shell element, and the fuel element, whose cross-sectional form is a rectangle (76.2 mmx76.2 mm), is modeled with the solid element that has equivalent stiffness. As the mass of the fuel element in the analysis, the largest mass of standard type fuel element is used. JRR-3 standard silicide type fuel is assumed as a representative.

omission Changes of item number due to deletion of JRR-3 aluminide fuel same as above same as above Deletion of JRR-4 fuel Modification for proper description due to deletion of JRR-3 aluminide fuel same as above

Comparison Table of SAR for Type JRC-80Y-20T Before After note

4. Strength of the fuel elements 4.1 Strength of JRR-3 standard aluminide type fuel As the horizontal drop direction, the X-direction and the Y-direction as shown in (II)-

Fig.A.106 are considered. Since the pressure area of the fuel element to the X-direction is small as compared with that to the Y-direction, the drop to the X-direction is severer than that of the Y-direction. Consequently, this paragraph shows the compressive strength of the fuel side plate at the drop to the X-direction.

The compressive stress (c) due to the inertia force is given by the following equation; c = WgGH A

where, W

Mass of the fuel element = 8.0 (kg) g

Gravitational acceleration = 9.8 (m/sec2)

GH

Impact deceleration = 167 (g)

A

Cross-sectional area of the fuel side plate

= 4.8x770x2 = 7.39x103 (mm2)

Therefore, the following is given; c = 8.0x9.8x167 7.39x103

= 1.78 MPa In this case, since the yield stress of material A6061-T6 at the temperature of 240 is less than that of material AG3NE, the yield stress of material A6061-T6 is used.

Consequently, the allowable compressive stress (ac) of the fuel element at the temperature of 240 is as follows; ac = 1.5xy = 1.5x46 = 69 MPa Consequently, the safety factor (RF) and the safety margin (MS) are as follows; RF = 69 1.78 = 38 MS = 38 1 = 37 As the above result, the fuel side plate of the fuel element has the sufficient strength against the compressive stress due to 9 m horizontal drop (II)-Fig.A.106 Horizontal drop direction of JRR-3 standard aluminide type fuel Fig. omitted 4.2 Strength of JRR-3 standard silicide type fuel The horizontal drop direction of JRR-3 standard silicide type fuel is considered to the X-direction and the Y-direction as shown in (II)-Fig.A.107 in the same manner as JRR-3 standard aluminide type fuel. The drop to the X-direction is severe, since the pressure area to the X-direction is smaller than that to the Y-direction.

Accordingly, this paragraph describes the compressive strength against the fuel side plate at the drop to the X-direction.

omission 4.3 Strength of JRR-3 follower aluminide type fuel The horizontal drop direction of JRR-3 follower aluminide type fuel is considered to the X-direction and the Y-direction as shown in (II)-Fig.A.108 in the same manner as JRR-4.1 Strength of JRR-3 standard silicide type fuel The horizontal drop direction of JRR-3 standard silicide type fuel is considered to the X-direction and the Y-direction as shown in (II)-Fig.A.106. The drop to the X-direction is severe, since the pressure area to the X-direction is smaller than that to the Y-direction.

Accordingly, this paragraph describes the compressive strength against the fuel side plate at the drop to the X-direction.

omission Deletion of JRR-3 aluminide fuel Changes of item number and drawing number due to deletion of JRR-3 aluminide fuel Deletion of JRR-3 aluminide fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note 3 standard aluminide type fuel. The drop to the X-direction is severe, since the pressure area to the X-direction is smaller than that to the Y-direction.

Accordingly, this paragraph describes the compressive strength against the fuel side plate at the drop to the X-direction.

The compressive stress (c) due to the inertia force is given as follows; c = WgGH A

where, W

Mass of the fuel element = 5.2 (kg) g

Gravitational acceleration = 9.8 (m/sec2)

GH

Impact deceleration = 167 (g)

A

Cross-sectional area of the fuel side plate (mm2)

= 4.8x770x2 = 7.39x103 (mm2)

Therefore, the following is given; c = 5.2x9.8x167 7.39x103

= 1.16 MPa In this case, the yield stress (y) of material AG3NE at the temperature of 240 is less than that of material A6061-T6, therefore, the yield stress (y) of 46 MPa of material AG3NE is used. Consequently, the allowable compressive stress (ac) at the temperature of 240 is given as follows ;

ab = 1.5xy = 1.5x46 = 69 MPa Consequently, the safety factor (RF) and the safety margin (MS) are as follows; RF = 69 1.16 = 59 MS = 59 1 = 58 As the above result, the fuel side plate has the sufficient strength against the compressive stress due to 9 m horizontal drop.

(II)-Fig.A.108 Horizontal drop direction of JRR-3 follower aluminide type fuel Fig. omitted 4.4 Strength of JRR-3 follower silicide type fuel The horizontal drop direction of JRR-3 follower silicide type fuel is considered to the X-direction and the Y-direction as shown in (II)-Fig.A.109 in the same manner as JRR-3 standard aluminide type fuel. The drop to the X-direction is severe, since the pressure area to the X-direction is smaller than that to the Y-direction.

Accordingly, this paragraph describes the compressive strength against the fuel side plate at the drop to the X-direction.

No change in description below 4.5 Strength of JRR-3 MNU type fuel No change in description below 4.6 Strength of JRR-4 low enrichment silicide type fuel The horizontal drop direction of JRR-4 low enrichment type fuel is considered to the X-direction and the Y-direction as shown in (II)-Fig.A.112 in the same manner as JRR-3 4.2 Strength of JRR-3 follower silicide type fuel The inertia force due to the vertical drop acts on the fuel element in the same manner as JRR-3 standard silicide type fuel element. The horizontal drop direction of JRR-3 follower silicide type fuel element is considered to the X-direction and the Y-direction as shown in (II)-Fig.A.107 in the same manner as JRR-3 standard silicide type fuel element.

The drop to the X-direction is severe, since the pressure receiving area to the X-direction is smaller than that to the Y-direction in terms of compressive stress on fuel elements.

omission 4.3 Strength of JRR-3 MNU type fuel omission Changes of item number and drawing number due to deletion of JRR-3 aluminide fuel Modification for proper description due to deletion of

Comparison Table of SAR for Type JRC-80Y-20T Before After note standard aluminide type fuel. The drop to the X-direction is severe, since the pressure area to the X-direction is smaller than that to the Y-direction.

Accordingly, this paragraph describes the compressive strength against the fuel side plate to the horizontal drop of the X-direction.

The compressive stress (c) due to the inertia force is given by the following equation; c = WgGH A

where, W

Mass of the fuel element = 5.6 (kg) g

Gravitational acceleration = 9.8 (m/sec2)

GH

Impact deceleration = 167 (g)

A

Cross-sectional area of the fuel side plate (mm2)

= 4.8x630x2 = 6.04x103 (mm2)

Therefore, c = 5.6x9.8x167 6.04x103

= 1.52 MP In this case, the yield stress (y) of material A6061-T6 at the temperature of 240 is less than that of material AG3NE, therefore, the yield stress (y) of 46 MPa of material A6061-T6 is used.

Consequently, the allowable compressive stress (ac) at the temperature of 240 is given as follows ;

ac = 1.5xy = 1.5x46 = 69 MPa Consequently, the safety factor (RF) and the safety margin (MS) are given as follows; RF = 69 1.52 = 45 MS = 45 1 = 44 As the result, the fuel side plate has the sufficient strength against the compressive stress at the time of 9 m horizontal drop.

(II)-Fig.A.112 Horizontal drop direction of JRR-4 low enrichment silicide type fuel Fig. omitted 4.7 Strength of JRR-4 high enrichment instrumented fuel (HEU)

The horizontal drop direction of JRR-4 low enrichment type fuel is considered to the X-direction and the Y-direction as shown in (II)-Fig.A.113 in the same manner as JRR-3 standard aluminide type fuel. The drop to the X-direction is severe, since the pressure area to the X-direction is smaller than that to the Y-direction.

Accordingly, this paragraph describes the compressive strength against the fuel side plate to the horizontal drop of the X-direction.

The compressive stress (c) due to the inertia force is given by the following equation; c = WgGH A

where, W

Mass of the fuel element = 6.0 (kg) g

Gravitational acceleration = 9.8 (m/sec2)

GH

Impact deceleration = 167 (g)

JRR-3 aluminide fuel Deletion of JRR-4 fuel Deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note A

Cross-sectional area of the fuel side plate (mm2)

= 4.8x630x2 = 6.04x103 (mm2)

Therefore, c = 6.0x9.8x167 6.04x103

= 1.63 MP In this case, the yield stress (y) of material A6061-T6 at the temperature of 240 is less than that of material AG3NE, therefore, the yield stress (y) of 46 MPa of material A6061-T6 is used.

Consequently, the allowable compressive stress (ac) at the temperature of 240 is given as follows ;

ac = 1.5xy = 1.5x46 = 69 MPa Consequently, the safety factor (RF) and the safety margin (MS) are given as follows; RF = 69 1.63 = 42 MS = 42 1 = 41 As the result, the fuel side plate has the sufficient strength against the compressive stress at the time of 9 m horizontal drop.

(II)-Fig.A.113 Horizontal drop direction of JRR-4 high enrichment instrumented fuel (HEU)

Fig. omitted

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission A.6.3.3 Comparison of allowable stress omission 4.

Thermal expansion of the shell and the basket generated during the fire accident a)

Clearance between the shell and the basket generated during the fire accident This paragraph shows the examination of JRR-3 standard aluminide fuel contained for box type fuel in which the heat generated by the contents is maximized and the clearance between the shell and the basket is minimized.

omission (II)-Fig.A.128 Temperature history of the basket for box type fuel contained JRR-3 standard aluminide type fuel Fig. omitted omission A.6.4 Water immersion It has been confirmed that permanent deformation does not occur and containment can be maintained at the water depth of 200 m (2.0MPa).

Therefore the packaging has enough strength and containment capacity at the water depth of 15m because the condition of immersion at the water depth of 200m is more severe than at the water depth of 15m.

omission A.6.5 Summary and evaluation of the results This paragraph shows summary and the evaluation of the results under accident condition of transport in accordance with each test item. (II)-Table A.20 shows the summary of the results of the structural analysis.

1. Drop test (9 m drop)

This item shows the summary of the result of the test.

(1)

In the packaging, the gasket portion of the lid flange is the severest part, where is still in the state of the elasticity in any drop attitude, and the stress of the lid bolt is restored after drop to the initial fastening stress. Therefore, the containment at the contact surface of the lid and the body is maintained and the shielding performance is not lost.

(Refer to,, and of (II )-Table A.20.)

(2)

In any drop attitude, the stress generated on the lid bolt is less than the yield stress of the material and the initial fastening stress is maintained after drop.

(Refer to,, and of (II )-Table A.20.)

(3)

Among the baskets, though a slight plastic deformation occurs partially on the basket for the box type fuel in the horizontal drop, in a criticality analysis that omission A.6.3.3 Comparison of allowable stress omission 4.

Thermal expansion of the shell and the basket generated during the fire accident a)

Clearance between the shell and the basket generated during the fire accident This section examines the case of a basket for box type fuel, assuming that it is loaded with fuel elements that have a higher calorific value than the contents and a minimum gap between the container body and the fuel basket (hereinafter referred to as fuel element A) in order to make the analysis more conservative.

omission (II)-Fig.A.124 Temperature history of the basket for box type fuel containing fuel element A Fig. omitted omission A.6.4 Water immersion Since this nuclear fuel package is a package containing nuclear fuel material, etc. with an amount of radioactivity exceeding 100,000 times A2 value (the ratio of the radioactivity of the contents to be loaded to the 100,000 times A2 value is approximately 4.5 so exceeds 1), it will be evaluated whether the containment device is not damaged for the 200m immersion test.

It has been confirmed that permanent deformation does not occur and containment can be maintained at the water depth of 200 m (2.0MPa).

Therefore the packaging has enough strength and containment capacity at the water depth of 15m because the condition of immersion at the water depth of 200m is more severe than at the water depth of 15m.

omission A.6.5 Summary and evaluation of the results This paragraph shows summary and the evaluation of the results under accident condition of transport in accordance with each test item. (II)-Table A.20 shows the summary of the results of the structural analysis.

1. Drop test (9 m drop)

This item shows the summary of the result of the test.

(1)

In the packaging, the gasket portion of the lid flange is the severest part, where is still in the state of the elasticity in any drop attitude, and the stress of the lid bolt is restored after drop to the initial fastening stress. Therefore, the containment at the contact surface of the lid and the body is maintained and the shielding performance is not lost.

(Refer to,, and of (II )-Table A.20.)

(2)

In any drop attitude, the stress generated on the lid bolt is less than the yield stress of the material and the initial fastening stress is maintained after drop.

(Refer to,, and of (II )-Table A.20.)

(3)

Among the baskets, though a slight plastic deformation occurs partially on the basket for the box type fuel in the horizontal drop, in a criticality analysis that Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Modification for proper description Changes of item number Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before After note considers that plastic deformation, no influence is brought about to the subcriticality as showing in appendix 3 of (II)-E.7.3.

(Refer to and of (II )-Table A.20.)

(4)

Among the fuel elements, the severest stress is generated in JRR-3 MNU type fuel contained in the basket during the vertical drop. Even in this case, it has the sufficient strength, since the safety factor against the stress is 1.2.

(Refer to,,, and of (II )-Table A.20.)

considers that plastic deformation, no influence is brought about to the subcriticality as showing in appendix 3 of (II)-E.7.3.

(Refer to and of (II )-Table A.20.)

(4)

Among the fuel elements, the severest stress is generated in JRR-3 MNU type fuel contained in the basket during the vertical drop. Even in this case, it has the sufficient strength, since the safety factor against the stress is 1.2.

(Refer to,, and of (II )-Table A.20.)

Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before After note 2.

Drop test-II (penetration test)

The severest stress is generated when the mild steel bar directly hits the protection cover.

Since the safety factor against the stress is 1.08, the mild steel bar will not penetrate the packaging.

Accordingly, the containment of the packaging can be maintained.

(Refer to and of (II )-Table A.20.)

3.

Thermal test The severest stress due to the maximum temperature gradient is generated on lid bolt.

The safety factor against the stress is 1.58.

Therefore, the packaging will not be damaged, and the containment can be maintained.

(Refer to of (II )-Table A.20.)

4.

Water immersion test The containment against the external pressure (0.15MP) equivalent to the water depth of 15 m can be sufficiently maintained.

(Refer to of (II )-Table A.20.)

2.

Drop test-II (penetration test)

The severest stress is generated when the mild steel bar directly hits the protection cover.

Since the safety factor against the stress is 1.08, the mild steel bar will not penetrate the packaging.

Accordingly, the containment of the packaging can be maintained.

(Refer to of (II )-Table A.20.)

3.

Thermal test The severest stress due to the maximum temperature gradient is generated on lid bolt.

The safety factor against the stress is 1.58.

Therefore, the packaging will not be damaged, and the containment can be maintained.

(Refer to and of (II )-Table A.20.)

4.

Water immersion test The containment against the external pressure (0.15MP) equivalent to the water depth of 15 m can be sufficiently maintained.

(Refer to of (II )-Table A.20.)

Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before After note Modification for proper description due to deletion of JRR-3 aluminide fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note Deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note Deletion of JRR-3 aluminide and JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission A.7 Enhanced water immersion test The integrity of containment of the package at the water depth of 200m should be maintained because this package has more than 10 thousand times radioactivity of A2 value.

omission A.8 Radioactive contents The contents of the package are seven kinds as follows.

1) JRR-3 standard aluminide type fuel

2) JRR-3 standard silicide type fuel

3) JRR-3 follower aluminide type fuel

4) JRR-3 follower silicide type fuel

5) JRR-3 MNU type fuel

6) JRR-4 low enrichment silicide type fuel

7) JRR-4 high enrichment instrumented fuel (HEU)

JRR-3 standard aluminide type fuel and JRR-3 follower aluminide type fuel are plate-type fuels where the fuel meats of uranium aluminum dispersion type are covered with aluminum alloy. JRR-3 standard silicide type fuel, JRR-3 follower silicide type fuel and JRR-4 low enrichment type fuel, are plate-type fuels where the fuel meats of uranium silicon aluminum dispersion type are covered with aluminum alloy. JRR-4 high enrichment instrumented fuel (HEU) is plate-type fuels where the fuel meats of uranium aluminum dispersion type are covered with aluminum alloy Also, JRR-3 MNU type fuel is a cylindrical fuel which is the metallic natural uranium covered with aluminum alloy.

JRR-3 standard aluminide type fuel, JRR-3 standard silicide type fuel, JRR-4 low enrichment silicide type fuel are cut off its top and bottom portions, JRR-4 high enrichment instrumented fuel are cut off its bottom portion, which do not contain uranium, before being loaded in the packaging.

The weight of those fuel elements are shown in paragraph (I)-C-5, (f), and the configurations are shown in (I)-Fig.D.1 through (I)-Fig.D.4.

A.10.4 Appendix-4

12. Cycle As shown below, all requirements specified in NE 3221-5 (d) in the reference [7] are satisfied.

Consequently, no analysis is required.

12.1 Cycle between atmospheric pressure and operating pressure Sa = 3 x 137 = 411 MPa (42 kgf/ mm2) omission A.7 Enhanced water immersion test Since this nuclear fuel package is a package containing nuclear fuel material, etc.

with an amount of radioactivity exceeding 100,000 times A2 value (the ratio of the radioactivity of the contents to be loaded to the 100,000 times A2 value is approximately 4.5 so exceeds 1), it will be evaluated whether the containment device is not damaged for the 200m immersion test.

omission A.8 Radioactive contents The contents of the package are three kinds as follows.

1) JRR-3 standard silicide type fuel

2) JRR-3 follower silicide type fuel

3) JRR-3 MNU type fuel JRR-3 standard silicide type fuel and JRR-3 follower silicide type fuel are plate-type fuels where the fuel meats of uranium silicon aluminum dispersion type are covered with aluminum alloy.

Also, JRR-3 MNU type fuel is a cylindrical fuel which is the metallic natural uranium covered with aluminum alloy.

JRR-3 standard silicide type fuel are cut off its top and bottom portions, which do not contain uranium, before being loaded in the packaging.

The weight of those fuel elements are shown in paragraph (I)-C-5, (f), and the configurations are shown in (I)-Fig.D.1 A.10.4 Appendix-4

12. Cycle As shown below, all requirements specified in NE 3221-5 (d) in the reference [7] are satisfied.

Consequently, no analysis is required.

12.1 Cycle between atmospheric pressure and operating pressure Sa = 3 x 137 = 411 MPa (42 kgf/ mm2)

Modification for proper description Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Deletion of JRR-4 Modification for proper description due to deletion of JRR-3 aluminide fuel and JRR-4

Comparison Table of SAR for Type JRC-80Y-20T Before After note According to Fig. I-9.2.1 in the reference [1], the number of cycles corresponding to the above Sa value is about 13,000 cycles.

Since the maximum predicted number of cycles used is 100 cycles, it follows that the requirements in the ASME Code are satisfied.

According to Fig. I-9.2.1 in the reference [1], the number of cycles corresponding to the above Sa value is about 13,000 cycles.

Since the maximum predicted number of cycles used is 300 cycles, it follows that the requirements in the ASME Code are satisfied.

Modification for proper description

Comparison Table of SAR for Type JRC-80Y-20T B. Thermal analysis B.1 Summary omission Each fuel element has different decay heat as shown in ()-Table B.4. The decay heat of the package has a maximum value 2.25 kW, when 40 pieces of JRR-3 standard aluminide type fuel are contained. These decay heats are calculated by using ORIGEN and ORIGEN-JR code.

The results of thermal analysis are summarized as follows.

(1)

Normal conditions of transport The maximum temperature of the outer surface of this package without insolation is 70 at ambient temperature of 38 when JRR-3 standard aluminide type fuels are contained. This value does not exceed 85, the standard value specified in the IAEA Regulations. The maximum temperature of the contents in the solar insolation is 223 when JRR-3 standard aluminide type fuels are contained and this is less than melting point of Aluminum alloy, 660, that is used for fuel cladding.

omission (2)

Accident conditions of transport The temperatures rise up to 298 at the fuel element, 384 at the outer surface of the package, 182 at the containment boundary of the drain valve when JRR-3 standard aluminide type fuels are contained under the accident conditions of transport.

omission

The maximum thermal stresses and thermal deformation of the package, occurring in the case of containing JRR-3 standard aluminide type fuel in the absence of insolation under normal conditions of transport and in the case of containing JRR-3 MNU type fuel under accident conditions of transport, are far less than the allowable values.

omission B. Thermal analysis B.1 Summary omission Each fuel element has different decay heat as shown in ()-Table B.4. In the evaluation, assuming a case where 40 assemblies of more conservative fuel element (hereafter referred to as fuel element A) than the contents are loaded so as to maximize the decay heat per nuclear fuel package, the value was set to 2.25 kW. These decay heats are calculated by using ORIGEN and ORIGEN-JR code.

The results of thermal analysis are summarized as follows.

(1)

Normal conditions of transport The maximum temperature of the outer surface of this package without insolation is 70 at ambient temperature of 38 when Fuel element As are contained. The maximum temperature of the contents in the solar insolation is 223 when Fuel element As are contained and this is less than melting point of Aluminum alloy, 660,

that is used for fuel cladding.

omission (2)

Accident conditions of transport The temperatures rise up to 298 at the fuel element, 384 at the outer surface of the package, 182 at the containment boundary of the drain valve when Fuel element As are contained under the accident conditions of transport.

omission

The maximum thermal stresses and thermal deformation of the package, occurring in the case of containing Fuel element A in the absence of insolation under normal conditions of transport and in the case of containing JRR-3 MNU type fuel under accident conditions of transport, are far less than the allowable values.

omission Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T B.4.1.1 Analytical model omission 3 Analytical model This packaging can contain seven kinds of fuel elements in three types of baskets.

The basket for box type fuel is used to transport JRR-3 standard aluminide type fuel, JRR-3 standard silicide type fuel, JRR-3 follower aluminide type fuel, JRR-3 follower silicide type fuel, JRR-4 low enrichment silicide type fuel and JRR-4 high enrichment instrumented type fuel (HEU). The basket for MNU type fuel is used to transport JRR-3 MNU type fuel. Namely, the package has two kinds of basket.

omission 3.1 Analytical model when the basket for box type fuel installed.

Six kinds of fuel elements, JRR-3 standard aluminide type fuel, JRR-3 standard silicide type fuel, JRR-3 follower aluminide type fuel, JRR-3 follower silicide type fuel, JRR-4 low enrichment silicide type fuel and JRR-4 high enrichment instrumented type fuel (HEU) are contained in the basket for box type fuel.

omission The fuel element used for the analytical model is JRR-3 standard aluminide type fuel that has the maximum decay heat as the fuel elements contained in the basket for box type fuel.

()-Fig.B.1 The general view of the analytical model of the package containing the basket for box type fuel (In case of containing JRR-3 standard aluminide type fuel)

()-Fig.B.2 The longitudinal sectional view of the analytical model containing the basket for box type fuel (In case of containing JRR-3 standard aluminide type fuel)

()-Fig.B.3 The radial sectional view of the analytical model containing the basket for box type fuel (In case of containing JRR-3 standard aluminide type fuel)

Fig. omitted omission 4.1 Heat transfer in the package when the basket for box type fuel installed.

In solid material, heat transfer takes place by conduction.

For air gap in the package, heat transfer takes place by either convection or B.4.1.1 Analytical model omission

()-Table B.4 Total decay heat 3 Analytical model This packaging can contain three kinds of fuel elements in three types of baskets.

The basket for box type fuel is used to transport JRR-3 standard silicide type fuel and JRR-3 follower silicide type fuel. The basket for MNU type fuel is used to transport JRR-3 MNU type fuel. Namely, the package has two kinds of basket.

omission 3.1 Analytical model when the basket for box type fuel installed Two kinds of fuel elements, JRR-3 standard silicide type fuel and JRR-3 follower silicide type fuel, are contained in the basket for box type fuel.

omission The fuel element used for the analytical model is Fuel element A that has the maximum decay heat as the fuel elements contained in the basket for box type fuel.

()-Fig.B.1 The general view of the analytical model of the package containing the basket for box type fuel (In case of containing Fuel element A)

No change of drawing

()-Fig.B.2 The longitudinal sectional view of the analytical model containing the basket for box type fuel (In case of containing Fuel element A)

No change of drawing

()-Fig.B.3 The radial sectional view of the analytical model containing the basket for box type fuel (In case of containing Fuel element A)

No change of drawing omission 4.1 Heat transfer in the package when the basket for box type fuel installed In solid material, heat transfer takes place by conduction. For air gap in the package, heat transfer takes place by either convection or conduction and by radiation. For instance, for the basket containing Fuel element A, the convection is dominant in the lodgement for neutron source near center axis and the air gap above the fuel, and at other Deletion of JRR-3 aluminide fuel and JRR-4 fuel Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T conduction and by radiation. For instance, for the basket containing JRR-3 standard aluminide type fuel, the convection is dominant in the lodgement for neutron source near center axis and the air gap above the fuel, and at other locations the conduction is superior to the convection.

locations the conduction is superior to the convection.

Comparison Table of SAR for Type JRC-80Y-20T omission B.4.2 Maximum temperature This paragraph shows about i) the evaluation in the absence of insolation and ii) the evaluation of the maximum temperature.

1 Temperature evaluation in the absence of solar insolation.

In this evaluation, the steady-state thermal analysis when the package is exposed to ambient temperature of 38 in the absence of solar insolation is performed. The results obtained for the two kinds basket are shown in ()-Table B.5.

The temperature distributions on the main parts of the analytical model of the basket for the box type fuel containing JRR-3 standard aluminide type fuel and the basket for MNU type fuel containing JRR-3 MNU type fuel are shown in ()-Fig.B.7, ()-Fig.B.8 and ()-Fig.B.9, through ()-Fig.B.10 respectively.

As a result, the maximum temperature at the outer surface of the packaging is 70 at the center of the body bottom plate in case of containing JRR-3 standard aluminide type fuel which has the maximum decay heat.

omission B.4.2 Maximum temperature 1 Temperature evaluation in the absence of solar insolation.

In this evaluation, the steady-state thermal analysis when the package is exposed to ambient temperature of 38 in the absence of solar insolation is performed.

The results obtained for the two kinds basket are shown in ()-Table B.5.

The temperature distributions on the main parts of the analytical model of the basket for the box type fuel containing Fuel element A and the basket for MNU type fuel containing JRR-3 MNU type fuel are shown in ()-Fig.B.7, ()-Fig.B.8 and ()-Fig.B.9, through ()-

Fig.B.10 respectively.

As a result, the maximum temperature at the outer surface of the packaging is 70 at the center of the body bottom plate in case of containing Fuel element A which has the maximum decay heat.

Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T

()-Fig.B.7 Temperature in the absence of solar insolation in case of containing JRR-3 standard aluminide type fuel (Longitudinal cross section)

Fig. omitted

()-Fig.B.8 Temperature in the absence of solar insolation in case of containing JRR-3 standard aluminide type fuel (Radial cross section)

Fig. omitted

()-Fig.B.7 Temperature in the absence of solar insolation in case of containing Fuel element A (Longitudinal cross section)

No change of drawing

()-Fig.B.8 Temperature in the absence of solar insolation in case of containing Fuel element A (Radial cross section)

No change of drawing Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T 2 Evaluation of the maximum temperature omission The temperature distributions on the main parts of the analytical model of the basket for box type fuel containing JRR-3 standard aluminide type fuel are shown in ()-

Fig.B.11 and ()-Fig.B.12. The temperature distributions on the main parts of the analytical model of the basket for MNU type fuel containing JRR-3 MNU type fuel are shown in ()-Fig.B.13 and ()-Fig.B.14.

Summarizing the results, the maximum temperature of the package is 223 at the fuel cladding when JRR-3 standard aluminide type fuel which is the maximum decay heat are contained. The value is lower than the melting point of the fuel cladding made of aluminum alloy, 660.

omission

()-Fig.B.11 Temperature in solar insolation in case of containing JRR-3 standard aluminide type fuel (Longitudinal cross section)

Fig. omitted

()-Fig.B.12 Temperature in solar insolation in case of containing JRR-3 standard aluminide type fuel (Radial cross section)

Fig. omitted 2 Evaluation of the maximum temperature omission The temperature distributions on the main parts of the analytical model of the basket for box type fuel containing Fuel element As are shown in ()-Fig.B.11 and ()-Fig.B.12.

The temperature distributions on the main parts of the analytical model of the basket for MNU type fuel containing JRR-3 MNU type fuel are shown in ()-Fig.B.13 and ()-

Fig.B.14.

Summarizing the results, the maximum temperature of the package is 223 at the fuel cladding when Fuel element A which is the maximum decay heat are contained. The value is lower than the melting point of the fuel cladding made of aluminum alloy, 660.

omission

()-Fig.B.11 Temperature in solar insolation in case of containing Fuel element A (Longitudinal cross section)

No change of drawing

()-Fig.B.12 Temperature in solar insolation in case of containing Fuel element A (Radial cross section)

No change of drawing Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T omission B.4.4 Maximum internal pressure This package is sealed up after the verification that it is under the thermal equilibrium after containing the fuels.

Therefore, the maximum internal pressure of the package under the normal conditions of transport occurs by the internal temperature difference between at the time of sealing up under solar insolation. The details are shown in appendix paragraph ()-

B.6.3. The results obtained are shown in ()-Table B.6. The maximum internal pressure occurs when JRR-3 standard aluminide type fuels are contained, and is 0.0517 MPaG. This valve is much smaller than the test pressure of 0.98 MPaG (10 kgf /cm2G).

Therefore, it is assured that no problem of pressure rise exists under the normal conditions of transport.

B.4.5 Maximum thermal stress The maximum thermal stress of this package occur at the center of the body bottom plate under the normal conditions of transport in the absence of insolation when JRR-3 standard aluminide type fuels are contained, and the value of the stress is 59.4 MPa and the safety margin MS is 5.9.

The stress of the lid bolts (due to the initial fastning force + the maximum internal pressure + the thermal load) is 115 MPa, and the safety margin MS is 3.3. There is no problem of the containment for the contact surface between the body and the lid.

The maximum thermal expansion of the basket occurs when JRR-3 standard aluminide type fuels are contained in the same as the maximum thermal stress. In this case, the expansion of the basket in the longitudinal direction and the radial direction are 1.44 mm and 0.981 mm respectively. For the above values, the gap in the longitudinal direction is 4.526 mm.

B.4.6 Summary of the results and the evaluation under normal conditions of transport

1. Surface temperature in the absence of solar insolation The maximum surface temperature of the package in the absence of solar insolation is 70 when the packaging contains JRR-3 standard aluminide type fuels generating the maximum decay heat. The value is below 85 specified in the technical standard.

2. Maximum temperature (Melting)

The maximum temperature of each location of the package under the normal conditions of transport is 223 at the fuel element when the packaging contains JRR-3 standard aluminide type fuel. This value is much below 660, melting point of the fuel cladding, made of aluminum alloy, and also much below 1400, melting point of main parts of the packaging made of stainless steel.

3 Maximum internal pressure, maximum thermal stress and thermal expansion (deformation)

) The maximum internal pressure of this package under normal conditions of transports is 0.0517 MPaG when the packaging contains JRR-3 standard aluminide type omission B.4.4 Maximum internal pressure This nuclear fuel package is sealed up after the verification that it is under the thermal equilibrium after containing the fuels.

Therefore, the maximum internal pressure of the package under the normal conditions of transport occurs by the internal temperature difference between at the time of sealing up under solar insolation. The details are shown in appendix paragraph ()-B.6.3. The results obtained are shown in ()-Table B.6. The maximum internal pressure occurs when Fuel element As are contained, and is 0.0517 MPaG. In addition, even when considering the ambient temperature change expected during transportation (from -40 to 38), it is 0.0460 MPaG. This value is sufficiently small compared to the pressure proof test pressure of 0.98 MPaG (10 kgf/cm2G) or higher, then there is no problem due to pressure increase under general test conditions for this nuclear fuel package, and there is no risk of cracks or failures to the package.

B.4.5 Maximum thermal stress The maximum thermal stress of this package occur at the center of the body bottom plate under the normal conditions of transport in the absence of insolation when Fuel element As are contained, and the value of the stress is 59.4 MPa and the safety margin MS is 5.9.

The stress of the lid bolts (due to the initial fastning force + the maximum internal pressure + the thermal load) is 115 MPa, and the safety margin MS is 3.3. There is no problem of the containment for the contact surface between the body and the lid.

The maximum thermal expansion of the basket occurs when Fuel element As are contained in the same as the maximum thermal stress. In this case, the expansion of the basket in the longitudinal direction and the radial direction are 1.44 mm and 0.981 mm respectively. For the above values, the gap in the longitudinal direction is 4.526 mm.

B.4.6 Summary of the results and the evaluation under normal conditions of transport

1. Surface temperature in the absence of solar insolation The maximum surface temperature of the package in the absence of solar insolation is 70 when the packaging contains Fuel element A generating the maximum decay heat. The value is below 85 specified in the technical standard.

2. Maximum temperature (Melting)

The maximum temperature of each location of the package under the normal conditions of transport is 223 at the fuel element when the packaging contains Fuel element A. This value is much below 660, melting point of the fuel cladding, made of aluminum alloy, and also much below 1400, melting point of main parts of the packaging made of stainless steel.

Modification for proper description due to deletion of JRR-3 aluminide fuel Modification based on the evaluation in "Consideration of aging of Nuclear Fuel package" due to revision of the regulations Modification for proper description Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T fuel. This value is much below the hydro test pressure of 0.98 MPaG (10 kgf/cm2G).

3 Maximum internal pressure, maximum thermal stress and thermal expansion (deformation)

) The maximum internal pressure of this package under normal conditions of transports is 0.0517 MPaG when the packaging contains Fuel element A. This value is much below the hydro test pressure of 0.98 MPaG (10 kgf/cm2G).

Same as above

Comparison Table of SAR for Type JRC-80Y-20T omission B.5.1.1 Analytical model omission

3. Analytical model The analytical model is the same model as that used under normal conditions of transport.

The thermal analysis under this conditions is performed by using two kinds of analytical models, namely, the analytical models of the basket for box type fuel containing JRR-3 standard aluminide type fuels and the basket for MNU type fuel containing JRR-3 MNU type fuels.

JRR-3 standard aluminide type fuel, which is the maximum decay heat among seven kinds of fuel elements and where the maximum temperature of the package occurs under the normal conditions of transports, is used in the evaluation of the temperature distribution and the maximum internal pressure. JRR-3 standard aluminide type fuel and JRR-3 MNU type fuel are used in the evaluation of the maximum thermal stress.

omission B.5.3 Temperature of the package The result of analysis under this condition, the values obtained for each portion of the package are shown in ()-Table B.8 collectively. Also the temperature history of the main portion (refer to ()-Fig.B.15) is shown in ()-Fig.B.16 through ()-Fig.B.18.

As a result of the analysis, the maximum temperature of each portion of the package occurs when the packaging contains JRR-3 standard aluminide type fuel generating the maximum decay heat. In this case, the maximum temperature of each portion of the package is 298 in the fuel, 216 in the basket, 783 in the edge of the fins, 384 in the outer surface of the body bottom plate and 182 in the packing of the containment boundary, drain valve.

B.5.4 Maximum internal pressure The maximum internal pressure is considered in the same way as that of the normal conditions of transport. The calculation method of the maximum internal pressure is shown in paragraph ()-B.6.6. of the appendix. The maximum internal pressure occurs when the packaging contains JRR-3 standard aluminide type fuels, which is 0.0747 MPaG. The result obtained is shown in ()-Table B.8 collectively.

omission B.5.1.1 Analytical model omission

3. Analytical model The analytical model is the same model as that used under normal conditions of transport.

The thermal analysis under this conditions is performed by using two kinds of analytical models, namely, the analytical models of the basket for box type fuel containing Fuel element As and the basket for MNU type fuel containing JRR-3 MNU type fuels.

As a fuel element to be analyzed in the evaluation of temperature distribution and maximum internal pressure, the JRR-3 standard silicide type fuel element has the largest decay heat among the three types, but more conservatively, we consider a fuel element A, which has the highest temperature distribution in the maximum temperature evaluation under normal conditions of transport. Fuel element A and JRR-3 MNU type fuel are used in the evaluation of the maximum thermal stress.

omission B.5.3 Temperature of the package The result of analysis under this condition, the values obtained for each portion of the package are shown in ()-Table B.8 collectively. Also the temperature history of the main portion (refer to ()-Fig.B.15) of the analytical model of the basket for box type fuel containing Fuel element A is shown in ()-Fig.B.16 through ()-Fig.B.18.

As a result of the analysis, the maximum temperature of each portion of the package occurs when the packaging contains Fuel element A generating the maximum decay heat.

In this case, the maximum temperature of each portion of the package is 298 in the fuel, 216 in the basket, 783 in the edge of the fins, 384 in the outer surface of the body bottom plate and 182 in the packing of the containment boundary, drain valve.

B.5.4 Maximum internal pressure The maximum internal pressure is considered in the same way as that of the normal conditions of transport. The calculation method of the maximum internal pressure is shown in paragraph ()-B.6.6. of the appendix. The maximum internal pressure occurs when the packaging contains Fuel element As, which is 0.0747 MPaG. The result obtained is shown in ()-Table B.8 collectively.

Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T omission

()-Fig.B.16 Temperature history in case of containing JRR-3 standard aluminide type fuel Fig. omitted

()-Fig.B.16 Temperature history in case of containing JRR-3 standard aluminide type fuel Fig. omitted omission B.5.5 Maximum thermal stress omission In the basket for box type fuel, which has the maximum thermal stress under the accident conditions of transport, the safety margin for the heat stress is 0.45 even if the conservative assumption is applied. The minimum gap of 0.152 mm between the shell body and the basket under the accident conditions of transport is caused in 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months after fire breakout when the packaging contains JRR-3 standard aluminide type fuels, which have the maximum heat decay. Therefore, the basket is not restrained.

Therefore, there is no problem with this package against the maximum thermal stress and the maximum thermal expansion.

omission

()-Fig.B.16 Temperature history in case of containing Fuel element A No change of drawing

()-Fig.B.17 Temperature history in case of containing Fuel element A No change of drawing

()-Fig.B.18 Temperature history in case of containing Fuel element A No change of drawing omission B.5.5 Maximum thermal stress omission In the basket for box type fuel, which has the maximum thermal stress under the accident conditions of transport, the safety margin for the heat stress is 0.45 even if the conservative assumption is applied. The minimum gap of 0.152 mm between the shell body and the basket under the accident conditions of transport is caused in 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months after fire breakout when the packaging contains Fuel element As, which have the maximum heat decay. Therefore, the basket is not restrained.

Therefore, there is no problem with this nuclear fuel package against the maximum thermal stress and the maximum thermal expansion.

Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T B.5.6 Summary of the result and the evaluation under the accident conditions of transport omission

2. Maximum temperature (Melting)

The maximum temperature of each location of the package under the accident conditions of transport occurs when the packaging contains JRR-3 standard aluminide type fuel. The value is 298 at the fuel element. It is much below 660, melting point of the fuel cladding made of aluminum alloy. Also, the maximum temperature of the packaging occurs at the top of the fin made of stainless steel which is the main material of the packaging, and the value is 783, which is much below 1400, melting point of the stainless steel.

omission B.6.1 Appendix-1 Details relating to heat transfer in the package omission 1.1 Heat transfer in the fuel elements The standard type fuel element and the follower type fuel element have the sectional configuration, where more than ten thin fuel plates are put between two fuel side plates, and the heat transfer is not constant in circumference. The evaluation of the heat transfer in JRR-3 standard aluminide type fuel, which has the maximum decay heat, is performed First of all, the heat transfer of the air layer in the fuel element is examined.

The thin air layers exist in the space of fuel plates in the fuel element. Therefore, it is examined which is dominant in the air layer, convection or conduction. ()-Fig.B.6.2 shows the position of JRR-3 standard aluminide type fuel in the basket.

omission

()-Fig.B.6.2 JRR-3 standard aluminide type fuel in the basket Fig. omitted omission Secondly, the heat transfer in the fuel is examined.

It is examined whether the heat transfer of the decay heat to the basket is dominant in the direction of the fuel side plate,or the fuel plate. The condition putting JRR-3 standard aluminide type fuel in the basket is shown in ()-Fig.B.6.3.

omission

()-Fig.B.6.3 Direction of heat transfer in JRR-3 standard aluminide type fuel Fig. omitted omission 1.2 Air convection in the basket Which is dominant, either convection or conduction, in the air layer which exists in the basket except for the portion of the fuel element is examined by the same way as mentioned in the previous paragraph, and the heat transfer coefficient is calculated for the air layer which the convection is dominant.

As the representation of the examination, the results for JRR-3 standard aluminide B.5.6 Summary of the result and the evaluation under the accident conditions of transport omission

2. Maximum temperature (Melting)

The maximum temperature of each location of the package under the accident conditions of transport occurs when the packaging contains Fuel element A. The value is 298 at the fuel element. It is much below 660, melting point of the fuel cladding made of aluminum alloy. Also, the maximum temperature of the packaging occurs at the top of the fin made of stainless steel which is the main material of the packaging, and the value is 783, which is much below 1400, melting point of the stainless steel.

omission B.6.1 Appendix-1 Details relating to heat transfer in the package omission 1.1 Heat transfer in the fuel elements The standard type fuel element and the follower type fuel element have the sectional configuration, where more than ten thin fuel plates are put between two fuel side plates, and the heat transfer is not constant in circumference. The evaluation of the heat transfer in Fuel element A, which has the maximum decay heat, is performed First of all, the heat transfer of the air layer in the fuel element is examined.

The thin air layers exist in the space of fuel plates in the fuel element. Therefore, it is examined which is dominant in the air layer, convection or conduction. ()-Fig.B.6.2 shows the position of Fuel element A in the basket.

omission

()-Fig.B.6.2 Fuel element A in the basket no change of drawing omission Secondly, the heat transfer in the fuel is examined.

It is examined whether the heat transfer of the decay heat to the basket is dominant in the direction of the fuel side plate, or the fuel plate. The condition putting Fuel element A in the basket is shown in ()-Fig.B.6.3.

omission

)-Fig.B.6.3 Direction of heat transfer in Fuel element A no change of drawing omission 1.2 Air convection in the basket Which is dominant, either convection or conduction, in the air layer which exists in the basket except for the portion of the fuel element is examined by the same way as mentioned in the previous paragraph, and the heat transfer coefficient is calculated for the air layer which the convection is dominant.

As the representation of the examination, the results for Fuel element As are presented.

Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T type fuel are presented.

omission

()-Table B.6.1 Heat transfer coefficient of convection (In case of containing JRR-3 standard aluminide type fuel)

( Table omission omission 1.4 Heat transfer between the bottom end surface of the fuel element and the upper basket bottom plate A slight air layer exists between the fuel element and the upper basket bottom plate, which is evaluated as follows.

The heat transfer area is considered as only the sectional area of the fuel side plate.

The thickness of the air layer is considered as the gap that occurs when the fuel element is inclined to the utmost physically possible limit in the basket lodgement. For instance, in case of JRR-3 standard aluminide type fuel, the inclination angle is 0.487 ° and the maximum air layer thickness is 0.65 mm, and the mean value is 0.33 mm. Therefore, as the air layer thickness, the mean value (0.33 mm) is used.

omission B.6.3. Calculation of the maximum internal pressure under the normal conditions of transport This package is sealed after loading the fuel elements and achieving thermal equilibrium. Since there is only the air as the internal fluid in the package, the maximum internal pressure is obtained by calculating only the pressure increase due to the temperature increase between the temperature of the internal air in the packaging under the absence of insolation and that under solar insolation. For instance, when JRR-3 standard aluminide type fuel is contained, the minimum temperature of the internal air is considered to be equal to the temperature of the inner wall of the body in the results of Temperature evaluation in the absence of solar insolation, 56 (refer to

()-Fig. B.7).

omission B.6.5. Details with regard to calculation of the maximum temperature in the fuel element under the accident conditions of transport omission 1.

In case that the basket for box type fuel is contained (in case of containing JRR-3 standard aluminide type fuel in the packaging)

The temperature histories of each location under the accident conditions of transport are shown in ()-Fig.B.16 through ()-Fig.B.18, when the packaging contains JRR-3 standard aluminide type fuel.

omission B.6.6. Calculation of the maximum internal pressure under the accident conditions of transport The maximum internal pressure under the accident condition is evaluated in the omission

()-Table B.6.1 Heat transfer coefficient of convection (In case of containing Fuel element A) no change of drawing omission 1.4 Heat transfer between the bottom end surface of the fuel element and the upper basket bottom plate A slight air layer exists between the fuel element and the upper basket bottom plate, which is evaluated as follows.

The heat transfer area is considered as only the sectional area of the fuel side plate.

The thickness of the air layer is considered as the gap that occurs when the fuel element is inclined to the utmost physically possible limit in the basket lodgment. For instance, in case of Fuel element A, the inclination angle is 0.487 ° and the maximum air layer thickness is 0.65 mm, and the mean value is 0.33 mm. Therefore, as the air layer thickness, the mean value (0.33 mm) is used.

omission B.6.3. Calculation of the maximum internal pressure under the normal conditions of transport This package is sealed after loading the fuel elements and achieving thermal equilibrium. Since there is only the air as the internal fluid in the package, the maximum internal pressure is obtained by calculating only the pressure increase due to the temperature increase between the temperature of the internal air in the packaging under the absence of insolation and that under solar insolation. For instance, when Fuel element A is contained, the minimum temperature of the internal air is considered to be equal to the temperature of the inner wall of the body in the results of Temperature evaluation in the absence of solar insolation, 56 (refer to ()-Fig. B.7).

omission In addition, when the ambient temperature change expected during transportation

(-40°C to 38°C) is considered, it is assumed that the fuel is loaded in a -40°C environment and pressure is adjusted in a -40 °C environment. The internal temperature during pressure adjustment is determined by subtracting the ambient temperature change from the average temperature inside the container:

200 + 91 2

78 68 On the other hand, if the maximum temperature of internal air when subjected to the solar radiation heat in the same manner as in the maximum internal pressure calculation, is set to 223°C assuming it is equal to the maximum temperature of the fuel element in the maximum temperature evaluation results, then the internal pressure at this time will be Same as above Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Changes based on the re-evaluation of "Consideration of aging of nuclear fuel package" due to revision of the regulations

Comparison Table of SAR for Type JRC-80Y-20T same way as that under the normal conditions of transport.

The case that JRR-3 standard aluminide type fuels are contained is described. From Equation of paragraph B.6.4.

2 0.1013x1x496 341x2 0.1473 (MPa abs) 0.0460 (MPaG)

Since this is smaller than the calculated maximum internal pressure under normal conditions of transport, this will be covered by the above results, then even when considering the ambient temperature changes expected during transportation, this nuclear fuel package does not have a problem due to pressure increase under normal conditions of transport.

omission B.6.5. Details with regard to calculation of the maximum temperature in the fuel element under the accident conditions of transport omission 1.

In case that the basket for box type fuel is contained (in case of containing Fuel element A in the packaging)

The temperature histories of each location under the accident conditions of transport are shown in ()-Fig.B.16 through ()-Fig.B.18, when the packaging contains Fuel element A.

omission B.6.6. Calculation of the maximum internal pressure under the accident conditions of transport The maximum internal pressure under the accident condition is evaluated in the same way as that under the normal conditions of transport.

The case that Fuel element As are contained is described. From Equation of paragraph B.6.4.

Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note

. Containment analysis omission C.2.1 Containment system omission

4. Pressure and temperature As shown in ()-B Thermal Analysis, the pressure and temperature of this package containing forty JRR-3 standard aluminide type fuel becomes severest under normal and accident conditions of transport.

omission omission C.3 Normal conditions of transport Since the leakage of radioactive materials from this package under normal conditions of transport must be evaluated conservatively, JRR-3 standard aluminide type fuel that has maximum thermal condition is assumed to be contained in the package.

omission C.3.2 Pressurizing the containment system omission

) Temperature rise of the air temperature This package shall be sealed hermetically after loading fuel elements and achieving thermal equilibrium. At this time, the temperature of inner gases is 56

ºC (which shall be the lowest temperature of inner wall of packaging) when the packaging contains JRR-3 standard aluminide type fuel generating maximum decay heat. Under the normal condition of transport, the inner gas temperature is raised to 223 ºC (which shall be the maximum temperature of fuel element) due to solar heat. This temperature rise results in the pressurization of 0.0517 MPaG.

To sum up, the maximum inner pressure of the package shall be 0.0517 MPaG when the packaging contains JRR-3 standard aluminide type fuel.

omission C. Containment analysis omission C.2.1 Containment system omission

4. Pressure and temperature The pressure and temperature of this nuclear fuel package, as shown in (II)-B Thermal Analysis, will be the severest conditions under normal and accident conditions of transport when 40 fuel assemblies which have higher pressure and temperature (hereinafter referred to as fuel element A) than the fuel elements to be loaded are loaded to make the evaluation more conservative.

omission omission C.3 Normal conditions of transport Since the leakage of radioactive materials from this package under normal conditions of transport must be evaluated conservatively, fuel element A that has maximum thermal condition is assumed to be contained in the package.

omission C.3.2 Pressurizing the containment system omission

) Temperature rise of the air temperature This package shall be sealed hermetically after loading fuel elements and achieving thermal equilibrium. At this time, the temperature of inner gases is 56 ºC (which shall be the lowest temperature of inner wall of packaging) when the packaging contains fuel element A generating maximum decay heat. Under the normal condition of transport, the inner gas temperature is raised to 223 ºC (which shall be the maximum temperature of fuel element) due to solar heat. This temperature rise results in the pressurization of 0.0517 MPaG.

To sum up, the maximum inner pressure of the package shall be 0.0517 MPaG when the packaging contains fuel element A.

omission Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above Same as above Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T before after note C.4.1 Fission product gas omission

2. Discussion and results of case when the temperature rise of fuel element causes the fuel cladding to melt.

The maximum temperature (298 ) of the fuel element in fire condition occurs by the loading of JRR-3 standard aluminide type fuel element which generates maximum decay heat. Since the melting point of fuel cladding (aluminum alloy) is 660, the fuel cladding does not melt in this condition.

In manufacturing uranium aluminum dispersion type fuel such as JRR-3 standard aluminide type fuel and so on, the element is held at 450 for an hour and its integrity is confirmed by the blister test. According to the data concerning the temperature at which blisters are formed on irradiated fuel elements, the blister generating temperature is about 450 600 even for the maximum local fission density of 1.0x1021 fiss/cm3, which corresponds to JRR-3 standard aluminide type fuel. This is shown in ()-Fig.C.8.

C.4.1 Fission product gas omission

2. Discussion and results of case when the temperature rise of fuel element causes the fuel cladding to melt.

The maximum temperature (298 ) of the fuel element in fire condition occurs by the loading of fuel element A which generates maximum decay heat. Since the melting point of fuel cladding (aluminum alloy) is 660, the fuel cladding does not melt in this condition.

In manufacturing uranium aluminum dispersion type fuel such as fuel element A and so on, the element is held at 450 for an hour and its integrity is confirmed by the blister test. According to the data concerning the temperature at which blisters are formed on irradiated fuel elements, the blister generating temperature is about 450 600 even for the maximum local fission density of 1.0x1021 fiss/cm3, which corresponds to fuel element A. This is shown in ()-Fig.C.8.

Modification for proper description due to deletion of JRR-3 aluminide fuel Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before after note D. Shielding analysis omission D.2 Source specifications Seven kinds of the fuel elements are contained in the packaging. The burnup, the power density and the cooling time of each fuel element are shown in ()-Table D.1.

The specifications of the fuel element used for the shielding analysis are shown in ()-

Table D.2.

In the JRR-3 (standard aluminide type fuel and follower aluminide type fuel), one cycle of operation time is 35 days (27 days of operation and 8 days of shutdown). The packaging can contain 40 fuel elements of after burnup of 5 cycles at the maximum power 29.0MW. The minimum cooling time is 300 days. Four fuels cooled for 35 days longer thereafter are contained in the case of the standard aluminide type fuel, and two fuels cooled for 35 days longer thereafter are contained in the case of the follower aluminide type fuel respectively.

29.0MW 0MW 27 days 8 days 27 days 8 days 27 days 8 days 27 days 8 days 27 days Cooling time300+35(n-1) days Standard type: n=1, 2,,

10 Follower type: n=1, 2,,

20 Cycle omission In the JRR-4 (low enrichment silicide type fuel), the one cycle of operation time is 7 days (operation of 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months a day from Tuesday till Friday), and 42 cycles in a year. In the calculation, the one cycle of operation period is assumed to be 365 days (49 days of operation, 316 days of shutdown) by collecting the portion of 42 cycles. The packaging can contain 40 fuel elements of after burnup of 8 cycles at the maximum power 4.7355MW.

The minimum cooling time is 110 days.

4.7355MW 0MW 49 days 316days 49 days 316 days 49 days 49 days 316 days 49 days 316 days 49 days Cooling time 110days Cycle In the JRR-4 (high enrichment instrumented fuel (HEU)), the one cycle of operation time is 7 days (operation of 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months a day from Tuesday till Friday), and 42 cycles in a year. In the calculation, the one cycle of operation period is assumed to be 365 days (49 days of operation, 316 days of shutdown) by collecting the portion of 42 cycles. The packaging can contain 40 fuel elements of after burnup of 2 cycles at the maximum power 4.032MW. The minimum cooling time is 10000 days.

4.032MW 0MW 49 days 316days 49 days Cooling time 10000days Cycle omission D. Shielding analysis omission D.2 Source specifications Three kinds of the fuel elements are contained in the packaging. The burnup, the power density and the cooling time of each fuel element are shown in ()-Table D.1.

The specifications of the fuel element used for the shielding analysis are shown in ()-

Table D.2. In addition, in order to make the shielding analysis more conservative, the evaluation was performed assuming the case where the fuel elements with a higher source intensity (hereinafter referred to as fuel element B) than the fuel elements to be loaded are loaded.

omission In the Fuel element B, the one cycle of operation time is 7 days (operation of 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months a day from Tuesday till Friday), and 42 cycles in a year. In the calculation, the one cycle of operation period is assumed to be 365 days (49 days of operation, 316 days of shutdown) by collecting the portion of 42 cycles. The packaging can contain 40 fuel elements of after burnup of 8 cycles at the maximum power 4.7355MW. The minimum cooling time is 110 days.

4.7355MW 0MW 49 days 316days 49 days 316 days 49 days 49 days 316 days 49 days 316 days 49 days Cooling time 110days Cycle omission Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel Deletion of JRR-3 aluminide fuel Modification for proper description due to deletion of JRR-4 fuel Deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before after note omission omission Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before after note omission omission D.3.1 Analytical model (1) Basket for box type fuel This basket can be loaded with 6 kinds of fuel elements.

For the gamma shielding calculation, JRR-3 standard silicide type fuel is used in the analysis because it has the maximum gamma source intensity per unit length.

For the neutron shielding calculation, JRR-4 low enrichment silicide type fuel is used in the analysis because it has the maximum neutron source intensity per unit length.

omission omission omission D.3.1 Analytical model (1) Basket for box type fuel This basket can be loaded with 2 kinds of fuel elements.

For the gamma shielding calculation, JRR-3 standard silicide type fuel is used in the analysis because it has the maximum gamma source intensity per unit length.

And for the neutron shielding analysis, in order to make the evaluation more conservative, the case will be studied where the fuel element Bs, which are assumed to have a higher neutron source intensity per unit length than the contents, are loaded.

omission Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel Same as above Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before after note

()-Fig.D.2 Neutron shielding analytical model with basket for box type fuel (In case of containing JRR-4 low enrichment silicide type fuel)

Fig. omitted omission D.5.2 Results and evaluation The results are shown in ()-Table D.13 and ()-Table D.14 for each basket.

The maximum gamma dose equivalent rate of the package occurs when JRR-3 standard silicide type fuels which has the maximum source intensity per unit length are contained in the basket for box type fuel. The maximum neutron dose equivalent rate occurs when JRR-4 low enrichment silicide type fuel which has the maximum source intensity per unit length.

omission D.6.1 Appendix-1 Neutron yields due to (,n) reaction (omission Element Ni JRR-3 JRR-3 JRR-4 JRR-4 JRR-3 JRR-3 JRR-3 Standard aluminide type Standard silicide type Low enrichment silicide type High enrichment instrumented type (HEU)

Follower aluminide type Follower silicide type MNU type Aluminum Silicon Uranium 4.80x1022 0

5.67x1021 3.11x1022 8.23x1021 1.23x1022 3.80x1022 6.03x1021 9.05x1021 5.20x1022 0

1.55x1021 4.80x1022 0

5.67x1021 3.11x1022 8.23x1021 1.23x1022 0

0 4.96x1022 omission Alpha emitter Ymix JRR-3 JRR-3 JRR-4 JRR-4 JRR-3 JRR-3 JRR-3 Standard aluminide type Standard silicide type Low enrichment silicide type High enrichment instrumented type (HEU)

Follower aluminide type Follower silicide type MNU type Pu-238 Pu-239 Pu-240 Am-241 Cm-242 Cm-244 7.39x10-7 3.74x10-7 3.80x10-7 7.28x10-7 1.80x10-6 1.14x10-6 6.03x10-7 3.07x10-7 3.12x10-7 5.94x10-7 1.47x10-6 9.31x10-7 6.50x10-7 3.30x10-7 3.35x10-7 6.40x10-7 1.58x10-6 1.00x10-6 7.39x10-7 3.74x10-7 3.80x10-7 7.28x10-7 1.80x10-6 1.14x10-6 7.39x10-7 3.74x10-7 3.80x10-7 7.28x10-7 1.80x10-6 1.14x10-6 6.03x10-7 3.07x10-7 3.12x10-7 5.94x10-7 1.47x10-6 9.31x10-7 0.0 0.0 0.0 0.0 0.0 0.0

()-Fig.D.2 Neutron shielding analytical model with basket for box type fuel (In case of containing Fuel element B)

Fig. omitted omission D.5.2 Results and evaluation The results are shown in ()-Table D.13 and ()-Table D.14 for each basket.

The maximum gamma dose equivalent rate of the nuclear fuel package occurs when JRR-3 standard silicide type fuels which has the maximum source intensity per unit length are contained in the basket for box type fuel.

The maximum neutron dose equivalent rate occurs when fuel element B which has the maximum source intensity per unit length.

omission D.6.1 Appendix-1 Neutron yields due to (,n) reaction omission omission Modification for proper description due to deletion of JRR-4 fuel Same as above Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before after note Alpha emitter Q (Bq=/package)

JRR-3 JRR-3 JRR-4 JRR-4 JRR-3 JRR-3 JRR-3 Standard aluminide type Standard silicide type Low enrichment silicide type High enrichment instrumented type (HEU)

Follower aluminide type Follower silicide type MNU type Pu-238 Pu-239 Pu-240 Am-241 Cm-242 Cm-244 1.39x1012 2.72x1011 3.53x1011 5.93x1010 1.47x1011 1.49x1010 4.30x1012 4.80x1011 7.40x1011 1.99x1011 4.44x1011 8.09x1010 6.79x1011 1.51x1011 2.26x1011 3.48x1010 2.83x1012 5.24x109 5.83x109 1.67x109 4.80x108 2.43x108 6.59x104 1.08x104 8.76x1011 1.73x1011 2.24x1011 5.03x1010 5.68x1010 9.22x109 2.75x1012 3.07x1011 4.73x1011 1.27x1011 2.84x1011 5.17x1010 1.06x1011 2.60x1012 9.31x1011 1.63x1011 1.35x108 3.34x107 Alpha emitter QYmix (n/s/ package)

JRR-3 JRR-3 JRR-4 JRR-4 JRR-3 JRR-3 JRR-3 Standard aluminide type Standard silicide type Low enrichment silicide type High enrichment instrumented type (HEU)

Follower aluminide type Follower silicide type MNU type Pu-238 Pu-239 Pu-240 Am-241 Cm-242 Cm-244 1.03x106 1.02x105 1.34x105 4.32x104 2.65x105 1.70x104 2.59x106 1.48x105 2.31x105 1.18x105 6.51x105 7.53x104 4.41x105 4.99x104 7.58x104 2.23x104 4.47x106 5.25x103 4.31x103 6.25x102 1.82x102 1.77x102 1.19x10-1 1.23x10-2 6.47x105 6.47x104 8.51x104 3.66x104 1.02x105 1.05x104 1.66x106 9.43x104 1.48x105 7.55x104 4.16x105 4.81x104 0.0 0.0 0.0 0.0 0.0 0.0 Total 1.59x106 3.82x106 5.07x106 5.29x103 9.47x105 2.44x106 0.0 Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before After note E. Criticality analysis omission E.2.1 Contents Seven kinds of the fuel elements, which are shown in ()-Table E.1, are contained in the packaging. These fuel elements are contained in the two kinds of the baskets.

These two baskets with the fuel elements are analyzed individually.

The fuel specifications used for the criticality analysis are shown in ()-Table E.2.

With regard to the basket for box type fuel, the analyses are performed when JRR-3 standard aluminide type fuel, JRR-3 standard silicide type fuel, JRR-4 low enrichment silicide type fuel, JRR-4 high enrichment instrumented fuel (HEU), JRR-3 follower aluminide type fuel and JRR-3 follower silicide type fuel are contained in the packaging respectively. The maximum number of contained fuel elements is 40.

JRR-3 follower aluminide type fuel and JRR-3 follower silicide type fuel are loaded with adapters.

With regard to the basket for MNU type fuel, the analysis is performed when JRR-3 MNU type fuels are contained in the packaging. The maximum number of contained fuel elements is 160.

The extremities of JRR-3 standard aluminide and silicide type fuel, and JRR-4 low enrichment silicide type fuel, where no fuel meat exists, are cut off before these fuel elements are contained into the packaging.

The bottom extremities of JRR-4 high enrichment instrumented fuel (HEU), where no fuel meat exists, are cut off before these fuel elements are contained into the packaging.

The configurations of the fuel elements after cut off are as follows. JRR-3 standard aluminide type fuel and JRR-3 standard silicide type fuel element are 80 cm long and its extremities of 2.5 cm in length are fuel structural materials. JRR-4 low enrichment silicide type fuel is 66 cm long and its top extremity of 3.5 cm and bottom extremity of 2.5 cm in length are fuel structural materials. JRR-4 high enrichment instrumented fuel (HEU) is 84.0 cm long and its top extremity of 21.5 cm and bottom extremity of 2.5 cm in length are fuel structural materials.

JRR-3 follower aluminide type fuel and JRR-3 follower silicide type fuel are 88 cm long and its bottom extremity of 9.05 cm and its top extremity of 3.95 cm in length are fuel structural materials. JRR-3 MNU type fuel is 93.3 cm long and its bottom extremity of 2.8 cm and its top extremity of 2.17 cm in length are fuel structural materials.

E. Criticality analysis omission E.2.1 Contents Three kinds of the fuel elements, which are shown in ()-Table E.1, are contained in the packaging. These fuel elements are contained in the two kinds of the baskets.

These two baskets with the fuel elements are analyzed individually.

The fuel specifications used for the criticality analysis are shown in ()-Table E.2.

With regard to the basket for box type fuel, the analyses are performed when JRR-3 standard silicide type fuel and JRR-3 follower silicide type fuel are contained in the packaging respectively. The maximum number of contained fuel elements is 40.

JRR-3 follower silicide type fuel are loaded with adapters.

With regard to the basket for MNU type fuel, the analysis is performed when JRR-3 MNU type fuels are contained in the packaging. The maximum number of contained fuel elements is 160.

The extremities of JRR-3 standard silicide type fuel where no fuel meat exists, are cut off before these fuel elements are contained into the packaging.

The configurations of the fuel elements after cut off are as follows. JRR-3 standard silicide type fuel element are 80 cm long and its extremities of 2.5 cm in length are fuel structural materials.

JRR-3 follower silicide type fuel are 88 cm long and its bottom extremity of 9.05 cm and its top extremity of 3.95 cm in length are fuel structural materials. JRR-3 MNU type fuel is 93.3 cm long and its bottom extremity of 2.8 cm and its top extremity of 2.17 cm in length are fuel structural materials.

Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel Same as above Same as above

Comparison Table of SAR for Type JRC-80Y-20T Before After note Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission E.2.3 Neutron poison omission The size and the positions to the basket lodgement are confirmed before installation.

Therefore, the validity of the analytical models and the number densities shown in paragraph.E.3 is assured.

Cadmium wire is used as the neutron poison for JRR-3 standard silicide type fuel, and JRR-3 follower silicide type fuel, but this material is ignored in the analysis to evaluate conservatively.

omission E.2.3 Neutron poison omission The size and the positions to the basket lodgement are confirmed before installation. Therefore, the validity of the analytical models and the number densities shown in paragraph.E.3 is assured.

Furthermore, an evaluation will be made on the loss rate of 10B in a hypothetical case of receiving neutron irradiation from the contents for 100 years to show that the boral plates do not lose their efficacy.

10B absorbs thermal neutrons and produces 10B (n, )7Li reaction.

The neutron absorption loss rate of 10B is expressed by the following equation:

(Neutron absorption loss rate) = (Neutron irradiation dose) x (Absorption reaction cross-section of 10B)

For the neutron irradiation dose for 100 years, using the value for fuel element B, which has the maximum source intensity per unit length, as shown in (II)-Table D.4, it will be:

1.03x105x100x365x24x3600 3.25x1014 n/cm2 Here, considering Absorption cross-section of 10B: 3837 x 10-24 (cm2) Note 1

Then, 3.25x1014 x 3837x10-24x100 1.3x10-4 This means that the loss of 10B is negligible and the neutron absorbing ability of the boral plate will not be lost.

Cadmium wire is used as the neutron poison for JRR-3 standard silicide type fuel, and JRR-3 follower silicide type fuel, but this material is ignored in the analysis to evaluate conservatively.

Note 1: Radioisotope Pocket Data Book, 12th Edition (published by the Japan Radioisotope Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel Changes due to re-evaluation of "Consideration of aging of nuclear fuel package" due to revision of the regulations Reference description based on re-valuation

Comparison Table of SAR for Type JRC-80Y-20T Before After note Association)

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission E.3.1.2 Analytical model of packages in array omission (1) Basket for box type fuel The analytical models used in the criticality analysis are shown in ()-Fig.E.1 through

()-Fig.E.14. The model mainly consists of 3 parts.

omission position of the fuel elements The fuel elements are assumed to lean towards the center of the basket as shown in ()-Fig. E.2, E.6, E.9 and E.12. This assumption is conservative, since the critical size is assumed to be smaller.

()-Fig.E.1 Analytical model of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 standard aluminide or silicide type fuel]

Fig. omitted

()-Fig.E.2 Analytical model of containing the basket for box type fuel (Cross section of basket)[In case of containing JRR-3 standard aluminide or silicide type fuel]

Fig. omitted

()-Fig.E.3 Cross section of JRR-3 standard aluminide type fuel (Fig. omitted

()-Fig.E.4 Cross section of JRR-3 standard silicide type fuel Fig. omitted

()-Fig.E.5 Analytical model of containing the basket for box type fuel (Fig. omitted (Axial direction) [In case of containing JRR-4 low enrichment silicide type fuel]

(Fig. omitted)

()-Fig.E.6 Analytical model of containing the basket for box type fuel (Cross section of basket) [In case of containing JRR-4 low enrichment silicide type fuel]

(Fig. omitted)

()-Fig.E.7 Cross section of JRR-4 low enrichment silicide type fuel (Fig. omitted)

()-Fig.E.8 Analytical model of containing the basket for box type fuel (Axial direction) [In case of containing JRR-4 high enrichment instrumented fuel (HEU)]

(Fig. omitted)

()-Fig.E.9 Analytical model of containing the basket for box type fuel (Cross section of basket) [In case of containing JRR-4 high enrichment instrumented fuel (HEU)]

(Fig. omitted)

()-Fig.E.10 Cross section of JRR-4 high enrichment instrumented fuel (HEU)

(Fig. omitted)

()-Fig.E.11 Analytical model of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 follower aluminide or silicide type fuel]

(Fig. omitted)

()-Fig.E.12 Analytical model of containing the basket for box type fuel (Cross section of basket) [In case of containing JRR-3 follower aluminide or silicide type fuel]

omission E.3.1.2 Analytical model of packages in array omission (1) Basket for box type fuel The analytical models used in the criticality analysis are shown in ()-Fig.E.1 through

()-Fig.E.6. The model mainly consists of 3 parts.

omission position of the fuel elements The fuel elements are assumed to lean towards the center of the basket as shown in ()-Fig. E.2 and E.6. This assumption is conservative, since the critical size is assumed to be smaller.

()-Fig.E.1 Analytical model of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 standard silicide type fuel]

(No change of drawing)

()-Fig.E.2 Analytical model of containing the basket for box type fuel (Cross section of basket)[In case of containing JRR-3 standard silicide type fuel] (No change of drawing)

()-Fig.E.3 Cross section of JRR-3 standard silicide type fuel (No change of drawing)

()-Fig.E.4 Analytical model of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 follower aluminide or silicide type fuel]

(No change of drawing)

()-Fig.E.5 Analytical model of containing the basket for box type fuel (Cross section of basket) [In case of containing JRR-3 follower aluminide or silicide type fuel]

Changes for drawing number Changes due to deletion of JRR-3 aluminide fuel and JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note (Fig. omitted)

()-Fig.E.13 Cross section of JRR-3 follower aluminide type fuel (Fig. omitted)

()-Fig.E.14 Cross section of JRR-3 follower silicide type fuel (Fig. omitted)

(No change of drawing)

()-Fig.E.6 Cross section of JRR-3 follower silicide type fuel (No change of drawing)

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission E.3.2 Atomic number density in each region of analytical model The atomic number densities of the elements in each region used in the analytical models are shown in ()-Table E.3.

The boron content in the boral plate is conservatively assumed to be 75% of the minimum guarantee value, 12.4 wt%.

JRR-4 low enrichment silicide type fuel has fifteen fuel plates. Though the uranium content in 13 pieces of the inside fuel plate is greater than that in 2 pieces of the outside fuel plate, all of 15 pieces are conservatively assumed to be the inside fuel plate in the analysis.

omission omission E.3.2 Atomic number density in each region of analytical model The atomic number densities of the elements in each region used in the analytical models are shown in ()-Table E.3.

The boron content in the boral plate is conservatively assumed to be 75% of the minimum guarantee value, 12.4 wt%.

omission Changes due to deletion of JRR-4 fuel Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note E.4.4 Calculation results omission E.7.1 Appendix-1 Safety of the package under routine conditions of transport The criticality safety is examined for the routine conditions of transport. Under the routine conditions of transport, the analysis is performed when there is no water inside and outside the packaging.

Basket for box type fuel The analytical models are the same as those shown in ()-Fig.E.1 through ()-

Fig.E.14. The density of the space region shown in ()-Table E.3 is assumed to be that of air. The KENO-Va code is used for the analysis.

The results of the analysis are as follows; JRR-3 standard aluminide type fuel (in case of containing 40 fuel elements) keff

= 0.124

= 0.0002 keff + 3

= 0.124 JRR-3 standard silicide type fuel (in case of containing 40 fuel elements)

E.4.4 Calculation results omission E.7.1 Appendix-1 Safety of the package under routine conditions of transport The criticality safety is examined for the routine conditions of transport. Under the routine conditions of transport, the analysis is performed when there is no water inside and outside the packaging.

Basket for box type fuel The analytical models are the same as those shown in ()-Fig.E.1 through ()-

Fig.E.6. The density of the space region shown in ()-Table E.3 is assumed to be that of air. The KENO-Va code is used for the analysis.

The results of the analysis are as follows; JRR-3 standard silicide type fuel (in case of containing 40 fuel elements)

Modification for proper description due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel Changes for drawing number Changes due to deletion of JRR-3 aluminide fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note keff

= 0.175

= 0.0002 keff + 3

= 0.176 JRR-4 low enrichment silicide type fuel (in case of containing 40 fuel elements) keff

= 0.101

= 0.0002 keff + 3

= 0.101 JRR-4 high enrichment instrumented fuel (HEU) (in case of containing 40 fuel elements) keff

= 0.068

= 0.0001 keff + 3

= 0.069 JRR-3 follower aluminide type fuel (in case of containing 40 fuel elements) keff

= 0.084

= 0.0001 keff + 3

= 0.084 JRR-3 follower silicide type fuel (in case of containing 40 fuel elements) keff

= 0.124

= 0.0002 keff + 3

= 0.124 From the above results, it can be concluded that the criticality safety is sufficiently kept.

keff

= 0.175

= 0.0002 keff + 3

= 0.176 fuel element B (in case of containing 40 fuel elements) keff

= 0.101

= 0.0002 keff + 3

= 0.101 JRR-3 follower silicide type fuel (in case of containing 40 fuel elements) keff

= 0.124

= 0.0002 keff + 3

= 0.124 From the above results, it can be concluded that the criticality safety is sufficiently kept.

Changes due to deletion of JRR-3 aluminide fuel and JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission E.7.2 Appendix-2 Safety of the package during the loading of the fuel elements When the fuel element is being loaded in this packaging, the lid is opened and the fuel element is perfectly surrounded by the water. In this state, the criticality safety is examined for each fuel basket.

Basket for box type fuel The axial model of this basket is shown in ()-Fig.E.2.1 through ()-Fig.E.2.3.

The model of the fuel element region is the same as that shown in ()-Fig.E.2, E.6, E.9 and E.12. Also, the model of the cross section of each fuel element is the same as that shown in ()-Fig.E.3, E.4, E.7, E.10, E.13 and E.14, respectively. The density of each component is the same as that shown in ()-Table E.3.

The thickness of surrounding water region is assumed to be 30 cm around the package, assuming the lid closed. The KENO-Va code is used for the analysis.

The results of the analysis are as follows; JRR-3 standard aluminide type fuel (in case of containing 40 fuel elements) keff

= 0.736

= 0.0008 keff + 3

= 0.738 JRR-3 standard silicide type fuel (in case of containing 40 fuel elements) keff

= 0.868

= 0.0010 keff + 3

= 0.870 JRR-4 low enrichment silicide type fuel (in case of containing 40 fuel elements) keff

= 0.767

= 0.0009 keff + 3

= 0.769 JRR-4 high enrichment instrumented fuel (HEU) (in case of containing 40 fuel elements) keff

= 0.699

= 0.0009 keff + 3

= 0.702 JRR-3 follower aluminide type fuel (in case of containing 40 fuel elements) keff

= 0.573

= 0.0008 keff + 3

= 0.576 omission E.7.2 Appendix-2 Safety of the package during the loading of the fuel elements When the fuel element is being loaded in this packaging, the lid is opened and the fuel element is perfectly surrounded by the water. In this state, the criticality safety is examined for each fuel basket.

Basket for box type fuel The axial model of this basket is shown in ()-Fig.E.2.1 through ()-Fig.E.2.2.

The model of the fuel element region is the same as that shown in ()-Fig.E.2 ~E.5.

Also, the model of the cross section of each fuel element is the same as that shown in ()-

Fig.E.3 and E.6, respectively. The density of each component is the same as that shown in ()-Table E.3.

The thickness of surrounding water region is assumed to be 30 cm around the package, assuming the lid closed. The KENO-Va code is used for the analysis.

The results of the analysis are as follows; JRR-3 standard silicide type fuel (in case of containing 40 fuel elements) keff

= 0.868

= 0.0010 keff + 3

= 0.870 Changes due to deletion of JRR-3 aluminide fuel and JRR-4 fuel Changes due to deletion of JRR-3 aluminide fue Changes due to deletion of JRR-3 aluminide fuel and JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note JRR-3 follower silicide type fuel (in case of containing 40 fuel elements) keff

= 0.695

= 0.0008 keff + 3

= 0.697 From the above results, it can be concluded that the criticality safety is sufficiently kept.

omission

()-Fig.E.2.1 Analytical model in case of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 standard aluminide or silicide type fuel]

Fig omitted

()-Fig.E.2.2 Analytical model in case of containing the basket for box type fuel (Axial direction) [In case of containing JRR-4 low enrichment silicide type or high enrichment instrumented fuel]

Fig omitted

()-Fig.E.2.3 Analytical model in case of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 follower aluminide or silicide type fuel]

Fig omitted

()-Fig.E.2.4 Analytical model in case of containing the basket for MNU type fuel Fig omitted JRR-3 follower silicide type fuel (in case of containing 40 fuel elements) keff

= 0.695

= 0.0008 keff + 3

= 0.697 From the above results, it can be concluded that the criticality safety is sufficiently kept.

omission

()-Fig.E.2.1 Analytical model in case of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 standard silicide type fuel]

(No change of drawing)

()-Fig.E.2.2 Analytical model in case of containing the basket for box type fuel (Axial direction) [In case of containing JRR-3 follower silicide type fuel]

(No change of drawing)

()-Fig.E.2.4 Analytical model in case of containing the basket for MNU type fuel (No change of drawing)

Changes due to deletion of JRR-3 aluminide fuel and JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission E.7.3 Appendix-3 Safety of the package under accident conditions of transport The basket for box type fuel of the package has very small deformation at 9 m drop test under accident condition of transport. The criticality safety of the package under this condition is confirmed.

As shown in ()-A.9.2, the basket for box type fuel deforms 0.7 mm in maximum under 9 m drop test. Fig. ()-E.3.1 shows the maximum displacement after having 9 m drop test, and the criticality calculation model of the basket for box type fuel after 9 m drop test is shown in Fig. ()-E.3.2. The model is same with the model shown in Fig.

()-E.1 through E.14 except the deformation of the basket. Calculations were performed by KENO-Va.

The results of the analysis are as follows; JRR-3 standard aluminide type fuel (in case of containing 40 fuel elements) keff

= 0.737

= 0.0009 keff + 3

= 0.740 JRR-3 standard silicide type fuel (in case of containing 40 fuel elements) keff

= 0.869

= 0.0009 keff + 3

= 0.872 JRR-4 low enrichment silicide type fuel (in case of containing 40 fuel elements) keff

= 0.770

= 0.0009 keff + 3

= 0.772 JRR-4 high enrichment instrumented fuel (HEU) (in case of containing 40 fuel elements) keff

= 0.698

= 0.0009 keff + 3

= 0.700 JRR-3 follower aluminide type fuel (in case of containing 40 fuel elements) keff

= 0.575

= 0.0008 keff + 3

= 0.577 JRR-3 follower silicide type fuel (in case of containing 40 fuel elements) keff

= 0.697

= 0.0009 keff + 3

= 0.700 omission E.7.3 Appendix-3 Safety of the package under accident conditions of transport The basket for box type fuel of the package has very small deformation at 9 m drop test under accident condition of transport. The criticality safety of the package under this condition is confirmed.

As shown in ()-A.9.2, the basket for box type fuel deforms 0.7 mm in maximum under 9 m drop test. Fig. ()-E.3.1 shows the maximum displacement after having 9 m drop test, and the criticality calculation model of the basket for box type fuel after 9 m drop test is shown in Fig. ()-E.3.2. The model is same with the model shown in Fig. ()-E.1 through E.8 except the deformation of the basket. Calculations were performed by KENO-Va.

The results of the analysis are as follows; JRR-3 standard silicide type fuel (in case of containing 40 fuel elements) keff

= 0.869

= 0.0009 keff + 3

= 0.872 JRR-3 follower silicide type fuel (in case of containing 40 fuel elements) keff

= 0.697

= 0.0009 keff + 3

= 0.700 Changes due to deletion of JRR-3 aluminide fuel and JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T Before After note From the results above, it can be concluded that the criticality safety is sufficiently kept.

From the results above, it can be concluded that the criticality safety is sufficiently kept.

Comparison Table of SAR for Type JRC-80Y-20T Before After note omission E.7.4 Appendix-4 Investigation of the optimum water density in the criticality evaluation omission omission E.7.4 Appendix-4 Investigation of the optimum water density in the criticality evaluation omission Changes to the figure due to deletion of JRR-3 aluminide fuel and deletion of JRR-4 fuel

Comparison Table of SAR for Type JRC-80Y-20T before after note F Consideration of Aging of Nuclear Fuel Package This chapter describes the matters which are to be considered in the safety analysis in Chapter (II) with regard to aging of nuclear fuel package component materials during the planned period of use of the transport container.

F.1 Aging Factors to be Considered For the nuclear fuel package, based on the anticipated conditions of use as shown in (II)-Table F.1, possible aging factors to be considered for the component materials of the transport container are thermal degradation, degradation due to irradiation, degradation due to chemical changes, and fatigue due to repeated stresses during container storage, before shipment, and during transportation.

The period of use of this package is 70 years from the time of manufacture, the frequency of use is once per year, and the number of days required for transport per transportation is conservatively 365 days. Assuming the number of handling times per transportation is 100 times, the total number of planned lifting times throughout the planned period of use is 7,000 times (100 times x 70 years) (A.4.4).

II)-Table F.1 Conditions of use anticipated during the planned period of use Status contents Conditions of use In storage No Transport containers are stored indoors.

In order to confirm that the performance of the transport container is maintained, a periodic voluntary inspection based on "Chapter (III) Maintenance of transport containers and handling methods of nuclear fuel packages" described in the application for design approval of nuclear fuel packages (Appendix-1) is to be performed at least once a year.

Before transportation Yes Nuclear fuel packages are to be stored indoors within the controlled area of the facility for up to three months from the time the contents are loaded to the time they are transported.

Before shipment of the package, a pre-shipment inspection based on "Chapter (III) Maintenance of transport containers and handling methods of nuclear fuel packages" is to be conducted.

During transportation Yes The package is to be transported by transport vehicle or vessel.

The package is to be securely tied to the vehicle or vessel so that it can withstand the shock and vibration Newly added chapter

Comparison Table of SAR for Type JRC-80Y-20T EHIRUH

DIWHU

QRWH

expected during transportation.

The period of transportation is expected to be about 2 months.

After transportation No After transportation, a visual inspection is to be conducted in controlled area (indoor) of the facility to confirm the integrity of the transport container.

Transport containers are stored indoors.

F.2 Evaluation of Necessity of Considering Aging in Safety Analysis Based on the aging factors shown in F.1, the necessity of considering the aging of each component material of the nuclear fuel package was evaluated with regard to thermal, radiation, and chemical changes that are expected during the planned period of use.

Fatigue evaluation was also conducted for the lifting device, which is subjected to loads during handling, and for the sealing device, which is subjected to loads due to changes in internal pressure. The results of these evaluations are shown in (II)-Table F.2.

The component materials of this nuclear fuel package are shown in Chapter (I) C.

Transport container, 3. Materials. Among these materials, those for which aging is to be considered are listed below.

Stainless Steel

%RURQ&DUELGH Aluminum alloy (spacer Note that aging of O-rings is not considered because they are replaced with each transportation.

Also, that aging of contents is not consider because they changes with each transportation.

Thhe in Stainless Steel

%RURQ &DUELGH spacer Aluminum alloy (s

%RURQ &DUELGH A

Note that aging of O-rings is not considered because they are replaced with each transportation.

Also, that aging of contents is not consider because they changes with each transportation.

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Table F.2 Evaluation of necessity of considering aging in safety analysis (1/3)

Component material Aging factors Evaluation Stainless Steel Heat Although there may be mechanical property degradations such as creep, etc. (deformation) by high temperature embrittlement due to exposure to a high temperature environment, the results of thermal analysis indicate that the temperature near the fuel basket center axis is 200 (the highest temperature during transportation is 223 for the fuel elements) (B.4.2), which is below the temperature (425 or higher) (1) at which deformation due to creep, etc. may occur. Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Radiation Although there may be effects on the mechanical properties due to microstructural changes (embrittlement, etc.) caused by neutron irradiation, the maximum neutron irradiation dose during the period of use is 2.27 x 1014 n/cm2, which is less than the dose of 1016 n/cm2(1) that may cause microstructural changes (embrittlement, etc.). Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Chemical changes Although there may be effects of corrosion on material strength, embrittlement, etc., stainless steel is a material that forms a passive film on its surface and is not susceptible to corrosion. The depth of corrosion in air is estimated to be 1m (0.001mm)(2) per year with the maximum of 0.07mm during the period of use, which is a negligible amount of corrosion compared to the thicknesses of the component materials (310mm for a transport container body). Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Fatigue (1) Lifting device Assuming the frequency of handling of the lifting device is 100 times per year, the realistic assumed number of lifting times during the period of use will be 7,000 times. However, the number of lifting times in compliance with the technical criteria is conservatively assumed to be 10,000 times, and the repeat count of 10,000 times covers the assumed number of uses. Based on the above, fatigue is evaluated with the repeat count being conservatively set to confirm that fatigue failure does not occur (A.4.4).

Comparison Table of SAR for Type JRC-80Y-20T before after note (2) Sealing device Assuming the frequency of handling of sealing devices is 4 times per year, the repeat count in 70 years will be 280 times.

However, the repeat count in compliance with the technical criteria is conservatively assumed to be 300 times, and this repeat count of 300 times covers the assumed number of uses.

Based on the above, fatigue is evaluated with the repeat count being conservatively set to confirm that fatigue failure does not occur (A.5.1.4).

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Table F.2 Evaluation of necessity of considering aging in safety analysis (2/3)

Component material Aging factors Evaluation Boron Carbide Heat Although there may be a functional degradation to maintain subcriticality due to microstructural changes caused by exposure to a high temperature environment, the results of the thermal analysis indicate that the maximum temperature during transportation is 223 (B.4.2), which is below the temperature at which this material melts (2450) (B.2). Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Radiation Although there may be a functional degradation to maintain subcriticality due to loss of 10B caused by neutron irradiation, the neutron irradiation dose is 3.25 x 1014 n/cm2 assuming conservatively the period of use of 100 years, the loss of 10B is estimated to be about 0.00013% (E.2.3), which means that the loss of 10B due to neutron irradiation is negligible. Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Chemical changes Although there may be a functional degradation to maintain subcriticality due to corrosion, corrosion does not occur because it is in a sealed space within the basket dividers (stainless steel), and does not come in contact with the outside air. Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Comparison Table of SAR for Type JRC-80Y-20T before after note (II)-Table F.2 Evaluation of necessity of considering aging in safety analysis (3/3)

Component material Aging factors Evaluation Aluminum alloy Heat Although there may be effects on the heat transfer performance due to the microstructural changes caused by exposure to a high temperature environment, the results of the thermal analysis indicate that the maximum temperature during transportation is 223(B.4.2), which is below the temperature at which this material melts (660) (B.2). Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Radiation Although there may be effects on the heat transfer performance due to microstructural changes (embrittlement, etc.) caused by neutron irradiation, the maximum neutron irradiation dose during the period of use is 2.27 x 1014 n/cm2, which is less than the dose of 1021 n/cm2(1) that may cause microstructural changes (embrittlement, etc.). Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Chemical changes Although there may be effects on the heat transfer performance due to corrosion, aluminum alloys form an oxide film on its surface and are not susceptible to corrosion (3). In addition, it is put to use after confirming that there are no abnormalities in its appearance before shipment. Based on the above, there is no need to consider the effect of aging in confirming compliance with the technical criteria.

Comparison Table of SAR for Type JRC-80Y-20T before after note F.3 Aging Considerations in Safety Analysis As described in F.2, the necessity of considering aging effects was evaluated for the component materials of the nuclear fuel package. As a result of the evaluation of aging effects, with regard to the factors of heat, radiation, and chemical changes, under the conditions of use expected during the planned period of use, it was confirmed that there is no need to consider their effects in confirming compliance with the technical criteria. For the lifting device and sealing device, it is necessary to consider aging effects due to fatigue because of repeated stresses. As a result of the evaluations of fatigue of the lifting device and sealing device, considering the conservative repeat count expected during the period of use, it was confirmed that fatigue failure did not occur, and therefore, there was no effect on conformance to the technical criteria.

F.4 Appendix F.4.1 Appendix-1 References (1) Transportation Technology Advisory Board, "Measures to Ensure Safety of Post-Storage Transportation for Interim Storage of Spent Fuel" (2010).

(2) Nikkan Kogyo Shimbun, Ltd. "Stainless Steel Handbook" (1979).

(3) Sumitomo Light Metal Industries, Ltd., "Aluminum Handbook (3rd Edition)" (1985).

Comparison Table of SAR for Type JRC-80Y-20T Before after note Chapter IV : Maintenance conditions of transport packaging and handling method of package Chapter III : Maintenance conditions of transport packaging and handling method of package Changes due to deletion of the previous chapter

Comparison Table of SAR for Type JRC-80Y-20T Before after note Chapter IV : Maintenance conditions of transport packaging and handling method of package omission Chapter III : Maintenance conditions of transport packaging and handling method of package With regard to the maintenance of transport containers and the handling methods of nuclear fuel packages that conform to the safety design of nuclear fuel packages (including consideration of aging), based on the results of the safety analysis ((II)-A to F),

the pre-shipment inspection at each transportation to confirm the integrity of nuclear fuel packages and the periodic voluntary inspection to ensure the performance of transport containers for the planned period of use will be conducted. The details are shown below.

omission to F), the pre-shipment inspection at each transportation to confirm the integrity of nuclear fuel packages and the periodic voluntary inspection to ensure the performance Addition of items for evaluation

Comparison Table of SAR for Type JRC-80Y-20T Before after note omission omission

Comparison Table of SAR for Type JRC-80Y-20T Before after note Addition of items for inspection Addition of description of spacer

Comparison Table of SAR for Type JRC-80Y-20T Type JRC-80Y-20T nuclear fuel Transport Package Basic policy for the quality management Newly added chapters APP

Comparison Table of SAR for Type JRC-80Y-20T Basic policy for quality management This quality management system stipulates the requirements for quality assurance activities by reference to the Rules of Quality Assurance for Safety of Nuclear Power Plants (JEAC4111-2009).

A. Quality management system A.1 General requirements (1) An organization shall establish, document, implement, and maintain a quality management system for transportation, etc. An organization shall also continue to improve the effectiveness of this quality management system.

(2) An organization shall implement the following matters:

(a) Clarifying processes required for a quality management system and their application to an organization.

(b) Clarifying the order and correlation of the processes.

(c) Defining required judgment criteria and methods to ensure that both operation and management of the processes are effective.

(d) Ensuring that the resources and information required to operate and monitor the processes are available.

(e) Monitoring, measuring, and analyzing the processes. However, the measurement can be skipped when it is difficult to measure.

(f) For the processes, taking measures required to obtain results as planned and continue to improve them.

(g) Matching the processes and the organization with a quality management system.

(h) Promoting work based on the knowledge of social science and behavioral science.

A.2 Requirements for documentation A.2.1 General The quality management system documents shall be each item of the following:

(1) Quality policy and quality objective (2) Primary document (quality assurance program)

(3) Secondary document (documents required by primary documents and documents such as rules determined necessary by an organization)

(4) Tertiary documents (documents such as procedures and guides determined necessary by an organization other than primary documents and secondary documents)

(5) Records required by documents of (1) to (4)

Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T A.2.2 Quality assurance plan The Director General shall develop, review as necessary, and maintain a quality assurance plan that includes the followings:

(1) Matters related to planning, implementation, evaluation, and improvement of the quality management system (2) Scope of application of the quality management system (3) Established "documented procedures" for the quality management system or information that makes it possible to refer to them (4) A description of the interrelationships among the processes of the quality management system.

A.2.3 Document management A director general and a manager (research reactor accelerator administration manager.

The same shall apply hereinafter) shall define procedures for the document and record management to certainly implement the following matters:

(1) Managing documents required by a quality management system. However, although records are a kind of documents, they are managed in accordance with the requirements specified in A.3 Record management.

(2) Specifying the management required for the following activities:

(a) Approving documents prior to the issuance from the viewpoint of whether they are appropriate.

(b) Reviewing, renewing as necessary, and reapproving documents.

(c) Clarifying the identification of document changes and the identification state of the currently effective version, by a management ledger, etc.

(d) Ensuring that the appropriate version of the corresponding document is available when and where it is required, by a management ledger, etc.

(e) Ensuring that documents can be easily read and easily distinguishable.

(f) Clarifying external documents determined to be required for the quality management system planning and operation and ensuring that their distribution is managed by a management ledger, etc.

(g) Preventing an abolished document from being used by mistake. Also, identifying it appropriately when it is retained for a certain purpose.

A.3 Quality record management A director general and a manager shall define procedures for the document and record Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T management to certainly implement the following matters:

(1)Clarifying the target for creating records and maintaining them to provide evidence of conforming to requirements and effectively operating a quality management system.

(2)Making records easy to read, easily distinguishable, and retrievable.

(3)Specifying the management required for identification, storage, protection, retrieval, storage time, and disposal of records.

Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T B. An applicant's responsibilities B.1 Commitment A director general shall conduct the following matters as the top's commitment to construct and implement a quality management system and continue to improve its effectiveness:

(1) Making it public in an organization to observe laws and ordinances and regulatory requirements.

(2) Setting up a quality policy.

(3) Promoting activities for fostering nuclear safety.

(4) Ensuring that quality objectives are set up.

(5) Conducting a management review.

(6) Ensuring that resources are available.

B.2 Emphasis on nuclear safety A director general shall give top priority to nuclear safety, determine requirements for work, and ensure that they are met.

B.3 Quality policy and quality objective B.3.1 Quality record A director general shall certainly conduct the following matters concerning the quality policy related to transportation, etc.:

(1) Being appropriate in regard to Article 4 of the Act on the Japan Atomic Energy Agency, Independent Administrative Agency (Purpose of the agency.)

(2) Being appropriate in regard to the quality policy concerning nuclear safety specified by the chief director.

(3) Incorporating the commitment to conform to requirements and continue to improve the effectiveness of a quality management system.

(4) Giving the framework for set-up and review of quality objectives.

(5) Making them transmitted to and understood by the whole organization.

(6) Reviewing to maintain their adequacy.

B.3.2 Quality objective The Director General should establish manuals for the management of quality objectives to ensure that the followings are implemented.

(1) The Director General shall have the Director set quality objectives. Such quality objectives shall include those necessary to meet the requirements for the work, if any.

(2) The quality objectives shall be consistent with the quality policy and the degree of the Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T achievement of those objectives shall be judgeable.

B.4 Responsibility and authority B.4.1 Structure The quality assurance organization for work concerning transportation containers, etc. is shown in (c)-Fig. B.2.

B.4.2 Responsibility and authority The following persons have responsibility and authority in the matters described for each:

(1) Director general A director general integrates and promotes quality assurance activities for transportation, etc. carried out at the research institute.

(2) Person in charge of quality assurance control A person in charge of quality assurance control has the following responsibility and authority:

(a) To ensure that a process required for a quality management system is established, implemented, and maintained.

(b) To report to a director general on the quality management system's implementation status and whether improvements need to be made.

(c) To ensure that the consciousness of compliance with applicable laws and ordinances and nuclear safety is enhanced across the organization.

(3) Manager A manager integrates and promotes quality assurance activities for transportation, etc. in a department under his/her jurisdiction.

(4) Section chief A section chief conducts quality assurance activities for transportation, etc. in a department under his/her jurisdiction.

(5) Quality assurance promotion committee The quality assurance promotion committee reviews important matters for promoting quality assurance activities and for quality assurance activities in the research institute and matters inquired by a director general.

(6) Safety review committee for nuclear facilities and safety review committee for facilities used The safety review committee for nuclear facilities and the safety review committee for facilities used review important matters for promoting operational safety activities and for operational safety activities in the research institute and matters inquired by a director general.

Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T B.4.3 Internal communication An organization shall use meetings, business communication memorandums, etc. to ensure information exchange to allow better internal communication. It shall also ensure that the information about the effectiveness of a quality management system is exchanged.

B.5 Management review A director general shall define procedures for the management review to certainly implement the following matters:

B.5.1 General (1) For the work concerning transportation, etc., a director general shall conduct a management review at least once a year to confirm that a quality management system continues to function appropriately, validly, and effectively.

(2) In this review, the evaluation of opportunities for improving a quality management system and the evaluation of the necessity for changes of a quality management system including a quality policy shall be conducted.

(3) Records of the result of a management review shall be maintained.

B.5.2 Input to a management review A person in charge of quality assurance control shall incorporate the following matters in the input to a management review:

(a) Audit results (b) How outsiders view the achievement of nuclear safety (c) Implementation status of a process (including the achievement status of quality objectives) and inspection and test results (d) Implementation status of activities for fostering the nuclear safety culture (e) Status of compliance with applicable laws and ordinances (f) Status of preventative measures and corrective actions (g) Follow-up to results of previous management reviews (h) Changes which may affect a quality management system (i) Proposals for improvement B.5.3 Output from a management review A director general shall incorporate the decisions and measures on the following matters in the output from a management review:

Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T (a) Improvement of the effectiveness of a quality management system and its processes (b) Improvements required for work planning and implementation (c) Necessity for resources Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T (C)-Fig. B.2 Quality assurance organization concerning design, etc. of nuclear fuel package Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T C. Education and training A manager shall define procedures for the education and training management to certainly implement the following matters:

(1) To clarify the competence required for the personnel engaged in the work.

(2) To assign a person capable of carrying out the work, using the required education, training, skills, and experience as the basis of judgment.

(3) To carry out education and training or OJT, etc. so that personnel can have the required competence.

(4) To evaluate the effectiveness of conducted education and training, etc.

(5) To make personnel recognize the meaning and importance of their activities and how they can contribute to achieving quality objectives.

(6) To maintain records concerning education and training track records, skills, and experience.

D. Design management D.1 Design and development program (1) A manager shall define procedures for design and development management to clarify processes required for designing and developing a transportation container (including a prototype container).

(2) A section chief shall formulate and manage a design and development program in accordance with the procedures for design and development management.

(3) A section chief shall clarify the following matters in the design and development program:

(a) Stage of design and development (b) Review, verification, and validation suitable for each stage of design and development (c) Responsibility and authority for design and development (4) The design and development program shall incorporate the following matters and clearly indicate them to those who carry out design and development (employees, etc. and contractors):

(a) To clarify design/development requirements, such as applicable laws and ordinances, standards, and design/development conditions, persons in charge of the review, approval, etc., and required design analysis, design verification, etc., as design documents.

(b) To define procedures for selecting the components important for transportation containers functions and the construction method applied to them and evaluating the validity, etc., and evaluate them.

(c) To define procedures for selecting, documenting, and approving an appropriate Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T disposition method when a change (including a deviation) from design and development requirements arises.

(d) To assign those who have appropriate experience and knowledge to the design and development work and make the required information and means available.

(e) To allow the persons other than an original designer to evaluate design and development documents.

(5) A section chief shall clarify the following matters and operate and manage the interface between the organizations involved in design and development to ensure effective communication and clear assignment of responsibility. It shall incorporate the design interface with a section in charge of manufacturing transportation containers and a section in charge of maintaining transportation containers. The interface shall also be provided with contractors as necessary.

(a) Interface between organizations or between contractors (i) Clarifying the responsibility for the interface of design and development (ii) Clarifying methods for creating, reviewing, approving, issuing, distributing, and revising design documents on the interface of design and development and responsible organizations (b) Communication between organizations or between contractors (i) Clarifying methods to position, examine, and approve the information about design and development information communication (ii) Clarifying the interface between the organization carrying out design and development and the one related to each stage of procurement, manufacturing, and maintenance (or the external organization)

(6) A section chief shall renew the program formulated in accordance with the progress of design and development as appropriate.

D.2 Input to design and development (1) A section chief shall clarify the requirement-related input, reflect it in design and development, and keep and manage the records. The following matters shall be incorporated in the input:

(a) Requirements for the function and performance of a transportation container (including a prototype container)

(b) Requirements, such as applicable laws and ordinances (c) Requirements for a quality assurance program (d) Information obtained from a previous similar design when applicable (e) Other requirements essential for design and development Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T (2) A section chief shall clarify in writing and implement the method for review and approval to prevent inappropriate data use in clarifying design and development requirements.

(3) A section chief shall review the input's adequacy. It shall be noted that there is no omission, no ambiguity, and no incompatibility in requirements.

D.3 Output from design and development (1) A section chief shall present the output from design and development in the form of a drawing, a specification, a report, a check sheet, etc. to allow the verification comparing it with the input to design and development. In that case, it shall be ensured that the output from design and development is in the following states:

(a) The requirements given in the input to design and development are satisfied.

(b) The information suitable for performing procurement and work is provided.

(c) The characteristics of a transportation container essential for safe use and proper use are clarified.

(d) When a demonstration test and manufacturing of prototype containers are outsourced for the validation of design and development, the judgment of acceptance of the related inspection and test is incorporated, or it is referenced.

(2) A section chief shall approve the output from design and development before proceeding to the next stage.

D.4 Review of design and development (1) A section chief shall perform a systematic review as planned, aiming at the following matters at an appropriate design and development stage. In this review, those who have screening skills, such as experts in other departments, shall be included, as necessary.

(a) To evaluate whether design and development results can satisfy the requirements.

(b) To clarify problems and propose necessary measures.

(2) A section chief shall keep and manage review result records and disposition records if any disposition is required.

D.5 Design and development verification (1) A section chief shall perform a verification as planned to ensure that the output from design and development satisfies the requirements given in the input to design and development at an appropriate design and development stage, considering the following matters:

(a) Method of design and development verification (i) The verification for one or more designs and developments, such as a review of design Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T and development, alternative calculation, a demonstration test, and the comparison with previous similar designs, are performed as appropriate.

(ii) Design and development are verified by the persons other than an original designer.

(b) Alternative calculation The design and development requirements, the adequacy of a calculation code, etc. are confirmed as well as an original design.

(c) Demonstration test Tests, such as a verification test and a performance test, are carried out considering the structural material and the structural system of a transportation container, environmental conditions, etc.

(d) Comparison with previous similar designs and development The comparison with design and development requirements, a structural system, a calculation code, etc. for a comparison target is performed to confirm the validity of design and development.

(2) A section chief shall keep verification result records and disposition records if any disposition is required.

D.6 Validation of design and development (1) A section chief shall perform a validation as planned at an appropriate stage of design and development to ensure that the design documents as a result of design and development (including safety analysis reports) satisfy the requirements according to the designated use or the intended use. Whenever feasible, a validation shall be completed prior to delivering or providing design documents (including safety analysis reports).

(2) A section chief shall keep validation result records and disposition records if any disposition is required.

D.7 Change management of design and development (1) When changing design and development, an organization shall clarify the reasons for change, sections changed, changed contents, the existence of the influence due to the change, circumstances of the change, etc. before changing them, appropriately perform a review, verification, and validation, and approve the change before implementation in accordance with the procedure for design and development management.

(a) Changing design and development (i) Design and development are changed by the same management method of design and development as the one applied to the original design.

(ii) The influence of the change in design and development on the safety of a transportation Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T container (including components, etc.) and design documents (including safety analysis reports) and the validity are evaluated.

(b) Transmitting changes of design and development The information concerning design change is transmitted to related organizations in writing as specified by a design and development program.

(2) An organization shall keep change review result records and disposition records if any disposition is required.

Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T E. Manufacturing order of a transportation container E.1 Procurement management A director general shall define procedures for the procurement management to ensure the following matters:

E.1.1 Procurement process (1) An organization shall ensure that procured products, etc. comply with specified procurement requirements.

(2) The method and degree of management for suppliers, procured products, etc. shall be defined depending on the influence of procured products, etc. on nuclear safety.

(3) An organization shall evaluate and select a supplier, using the supplier's capability of supplying procured products, etc. in accordance with the organization's requirements as the basis of judgment, based on the criteria of selection, evaluation, and reevaluation defined in the procedure for procurement management.

(4) An organization shall keep evaluation result records and disposition records if any disposition is required by the evaluation.

(5) An organization shall define the method for obtaining the technical information concerning nuclear safety required for maintenance or operation after the procurement of procured products, etc. and the method for the necessary disposition when sharing them with other departments.

E.1.2 Procurement requirements (1) A section chief shall clarify the requirements for procured products, etc. and include the relevant items among the following when necessary:

(a) Requirements for the approval of a product, a procedure, a process, and equipment (b) Requirements for qualification confirmation of personnel (c) Requirements for a quality management system (d) Requirements for a nonconformity report and nonconformity disposition (e) Matters necessary for activities to foster a nuclear safety culture (f) Matters for information management (g) Other matters necessary for procured products, etc.

(2) An organization shall ensure that specified procurement requirements are valid before transmitting them to a supplier.

E.1.3 Verification of procured products (1) A section chief shall define and perform required inspections or other activities to ensure that procured products satisfy specified procurement requirements.

Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T (2) When a verification is performed at a supplier's facility, a section chief shall clarify the verification procedure and procured products release method (permission for shipment) in procurement requirements.

(3) When receiving procured products, an organization shall make a procured products supplier submit a document recording the conformity status to procurement requirements.

F. Handling and maintenance An organization shall plan and conduct the handling and maintenance management of a transportation container in accordance with the following:

(1) Considering the following matters, a manager shall define a procedure for handling transportation containers to prevent erroneous operation of and damage on a transportation container while handling a transportation container under his/her jurisdiction.

(a) Inspection of handling equipment and preventive measures against erroneous operation of and damage on a transportation container during handling (b) Handling conditions of a transportation container (c) The shipping in/out conditions and method of a transportation container from a storage facility (d) Person responsible for handling (2) A section chief shall clearly indicate requirements while handling a transportation container and reflect them in preventing erroneous operation of and damage on a transportation container in accordance with a procedure for handling transportation containers.

(3) Considering the following matters, a manager shall define a procedure for maintenance management of a transportation container to maintain the design performance of a transportation container under his/her jurisdiction.

(a) Requirements of laws and ordinances, design documents, authorized or licensed matters, etc.

(b) Inspection method and procedure for a transportation container (c) Damage prevention measures in storage (d) Setting up storage method and storage areas considering environmental conditions, etc.

(e) Person responsible for maintenance and storage (4) A section chief shall clarify requirements of applicable laws and ordinances/regulations, design documents, authorized or licensed matters, etc. and reflect them in maintenance management of transportation containers in accordance with a procedure for maintenance management of transportation containers.

Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T (5) A section chief shall clarify and manage persons responsible for work to those who perform maintenance or storage (employees, etc. and contractors).

(6) When maintenance work of transportation containers is outsourced, a section chief shall make a contractor submit management manuals clarifying the following matters and manage them after obtaining the manager's approval, as necessary.

(a) Requirements of laws and ordinances/regulations, etc.

(b) Persons responsible for approval, review, work instructions, etc. of rules, manuals, instructions, etc. required for management (7) Considering safety importance, etc., a section chief shall conduct witness confirmation and record confirmation in the maintenance inspection of transportation containers (including components).

G. Measurement, analysis and improvement G.1 General (1) Organization shall plan and implement the process for monitoring, measurement and improvement required for the following matters:

a) Verify conformity of requirements for the duties.

b) Ensure conformity of the quality management system.

c) Continuously improve effectiveness of the quality management system.

(2) This shall include statistical methods, applicable methods, and determination on the extent to which they are used.

G.2 Internal audit The Director General shall establish manuals for internal audits to ensure the following.

(1) The Director General shall conduct an internal audit at least once a year on transportation and other activities during the relevant fiscal year to verify whether the following items of the quality management system are fulfilled:

a) Whether the quality management system conforms to the plan of operations, the requirements of the quality assurance plan, and the quality management system requirements determined by the organization.

b) Whether the quality management plan has been effectively operated and maintained.

(2) The Director General should implement the internal program which specifies the following matters by taking into account the process to be the audited, its importance, the past audit results etc.

a) Criteria, scope and methods of audit b) Objectivity and fairness shall be ensured in selecting the auditor and implementing the Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T audit. Further, the auditor shall not audit his or her own duty.

(3) The manual for internal audits shall specify responsibilities and authorities (authority to order special internal audits) and requirements for planning and conducting audits, reporting results, and management of records.

(4) Records of audits and the results of them shall be retained.

(5) The person responsible for the audited area shall ensure that the necessary corrective and preventive actions are taken without delay to eliminate the nonconformity found and its cause. The follow-up shall include the verification of the actions taken and the report of the verification result.

G.3 Nonconformity control The Director General shall establish manuals for nonconformity management and corrective and preventive actions to ensure the followings:

(1) The organization shall ensure that nonconformities are identified and controlled to prevent them from being left unresolved. The manuals for nonconformity and corrective and preventive actions shall specify the controls over the handling of nonconformities and the responsibilities and authorities related thereto. The manual for internal audits shall specify the controls over, and the responsibility and authority for, the handling of nonconformities in quality assurance activities identified during internal audits.

(2) The organization shall take actions for nonconformity in either of the following ways:

a) Take action to remove detected nonconformity.

b) An authorized person may determine its use, release, or acceptance by special employment.

c) Take action to prevent its original intended use or application d) If a nonconformity is detected after delivery, the organization should take an appropriate action for the effects or possible effects of the nonconformity.

(3) The organization should maintain records of the nature of the nonconformity.

(4) When nonconformities are corrected, the organization shall reverify them to demonstrate conformance to the requirements.

(5) If a nonconformity is detected after delivery, the organization shall take appropriate action to address the effects or potential effects of the nonconformity.

G.4 Corrective actions The Director General shall establish manuals for nonconformity management and corrective and preventive actions, and for internal audits to ensure the followings.

(1) The organization shall take action to eliminate the causes of nonconformities to prevent Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T recurrence.

(2) Corrective actions shall be commensurate with the impact of the nonconformity that has been found.

(3) The following requirements shall be specified in the manuals for nonconformity management and corrective and preventive actions:

a) Confirmation of the nonconformity details b) Identification of the cause of nonconformity c) Evaluation of the necessity of the actions to certainly prevent nonconformity from occurring again d) Decision and performance of necessary actions e) Record of the results of the investigation and the corrective actions taken based on those results, when an investigation is conducted into corrective actions.

f) Review of activities performed in corrective actions (4) The following requirements shall be specified in the manual for internal audits:

a) Confirmation of the nonconformity details b) Identification of the cause of nonconformity c) Decision and performance of necessary actions d) Record of the results of actions that have been taken G.5 Preventative actions The Director General should establish manuals for nonconformity management and corrective and preventive actions, as well as manuals for horizontal deployment, to ensure the following.

(1) The organization should determine actions to eliminate the causes of possible nonconformities, including the acquisition and utilization of knowledge obtained through the implementation of safety activities and technical information obtained from inside and outside the institute, in order to prevent the occurrence of possible nonconformity. This utilization includes sharing the knowledge obtained through the implementation of nuclear safety and security-related activities with other organizations.

(2) Preventive actions shall be commensurate with the impact of possible problems.

(3) The organization shall specify requirements for the followings.

a) Identification of possible nonconformity and its cause b) Assessment of necessity of actions to ensure the prevention of non-conformity c) Determination and performance of necessary actions d) Record of the results of the investigation and the preventive actions taken based on those results, when an investigation is conducted into preventive actions.

Newly added chapters

Comparison Table of SAR for Type JRC-80Y-20T e) Review of activities performed in preventive actions Newly added chapters

v t e Letter Sequence Other
CAC:001794, (Approved, Closed) EPID:L-2023-DOT-0006, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance Questions 12 July 2023 30 May 2023 30 May 2023 2 June 2023 2 June 2023 30 May 2023 30 May 2023 2 June 2023 2 June 2023 30 May 2023 Answers 14 September 2023 27 June 2023 Results Approval, Approval Other: ML23058A279, ML23058A281, ML23058A284, ML23058A286, ML23058A287, ML23059A252, ML23067A441, ML23109A009, ML23115A080, ML23115A082, ML23115A086, ML23115A320, ML23160A084, ML23160A087, ML23310A163, ML23310A297, ML23310A299, ML23310A300, ML23310A301, ML23310A302, ML23310A303, ML23310A304, ML23310A305, ML23310A306, ML23331A675, ML23331A679, ML23331A681, ML23331A683, ML23331A684, ML23331A686, ML23331A688, ML23331A690, ML23331A693, ML23331A695, ML23331A698, ML23331A700, ML23331A702, ML23331A897, ML23331A898, ML23331A899, ML23331A900, ML23331A901, ML23331A902, ML23331A903, ML23331A904, ML23331A907, ML24022A254, ML24023A003, ML24023A321, ML24046A245... further results
MONTHYEAR2016-12-14 [Table View]

ML23115A082

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