Standard Review Plan for 

Spent Fuel Dry Storage Systems 

at a General License Facility 

Final Report 

Office of Nuclear Material Safety and Safeguards

NUREG-1536 

Revision 1

Standard Review Plan for 

Spent Fuel Dry Storage Systems 

at a General License Facility 

Final Report 

Manuscript Completed: July 2010 

Date Published: July 2010 

Office of Nuclear Material Safety and Safeguards 

NUREG-1536 

Revision 1 
 iii 

ABSTRACT 

The Standard Review Plan (SRP) for dry storage systems (DSS) provides guidance to the U.S. 

Nuclear Regulatory Commission (NRC) staff in the Division of Spent Fuel Storage and 

Transportation (SFST) for reviewing applications for a Certificate of Compliance (CoC) of a dry 

storage system (DSS) for use at a general license facility. This SRP is intended for use by the 

NRC staff. Its objectives are to: 

 • provide a basis that promotes a consistent regulatory review of an application for 

a DSS; 

 • promote quality and uniformity of these reviews across each technical discipline; 

 • present a basis for the review scope; 

 • identify acceptable approaches to meeting regulatory requirements; and 

 • develop an approach for review of each review procedure section of each 

chapter to assist the staff in prioritization of its review. 

Title 10 of the U.S. Code of Federal Regulations (CFR) Part 72 (10 CFR 72), Subpart B, 

specifies the information needed in a license application for the independent storage of spent 

nuclear fuel for a site specific application. Subparts A specifies the information needed in an 

application for a CoC for use at a general license facility. Regulatory Guide 3.61, Standard 

Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask, 

contains an outline of the information required by the staff. This SRP is divided into 14 chapters 

with appendices that reflect the standard application format. Regulatory requirements, staff 

positions, industry codes and standards, acceptance criteria, and other information are 

discussed. However, the format used herein has evolved and, in some instances, superseded 

Regulatory Guide 3.61 to better reflect current staff practice. 

In conjunction with the SRP, the SFST developed several Interim Staff Guidance (ISG) 

documents. An ISG addresses emergent review issues in a timely manner by staff and 

applicants. These ISGs were developed to address changes in requirements, reflect lessons 

learned and evolving technology, and document detailed technical positions. Current ISGs are 

available on the NRC website. Although Revision 1 of this SRP was revised to incorporate the 

applicable ISGs listed in Appendix C, other ISGs will continue to be developed as needed. This 

SRP will be revised periodically to reflect current guidance to the staff. 

The review procedures sections of each chapter of this SRP have been prioritized to assist the 

NRC staff in its review in an effort to increase efficiency. The method used to prioritize the 

Review Procedures sections is documented in Appendix B. The priority of each review 

procedure is shown in the applicable section of each chapter. 

Public Comments were solicited on this document and all public comments received are 

documented in Appendix D along with their resolution and associated changes to this SRP. 

Comments, errors or omissions, and suggestions for improvement should be sent to the 

Director, Division of Spent Fuel Storage and Transportation, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555-0001. 

 v 

TABLE OF CONTENTS 

ABSTRACT .................................................................................................................................. iii

TABLE OF CONTENTS ................................................................................................................ v

LIST OF FIGURES ..................................................................................................................... xii

LIST OF TABLES ........................................................................................................................ xiii

ACRONYMS AND ABBREVIATIONS ......................................................................................... xv

UNITS ........................................................................................................................................ xix

GLOSSARY ............................................................................................................................... xxi

INTRODUCTION .......................................................................................................................... 1

GENERAL INFORMATION EVALUATION ................................................................................ 1-1

1.1 Review Objective ................................................................................................ 1-1

1.2 Areas of Review ................................................................................................. 1-1

1.3 Regulatory Requirements ................................................................................... 1-1

1.4 Acceptance Criteria ............................................................................................ 1-2

1.4.1 DSS Description and Operational Features ...................................................... 1-2

1.4.2 Drawings ........................................................................................................... 1-3

1.4.3 DSS Contents .................................................................................................... 1-3

1.4.4 Quality Assurance ............................................................................................. 1-3

1.4.5 Consideration of 10 CFR Part 71 Requirements Regarding Transportation ..... 1-3

1.5 Review Procedures ............................................................................................ 1-3

1.5.1 DSS Description and Operational Features ...................................................... 1-4

1.5.2 Drawings (MEDIUM Priority) ............................................................................. 1-6

1.5.3 DSS Contents .................................................................................................... 1-6

1.5.4 Quality Assurance Program .............................................................................. 1-7

1.5.5 Consideration of 10 CFR Part 71 Requirements ............................................... 1-8

1.6 Evaluation Findings ............................................................................................ 1-8

2 PRINCIPAL DESIGN CRITERIA EVALUATION ................................................................... 2-1

2.1 Review Objective ................................................................................................ 2-1

2.2 Areas of Review ................................................................................................. 2-1

2.3 Regulatory Requirements ................................................................................... 2-1

2.4 Acceptance Criteria ............................................................................................ 2-2

2.4.1 SSCs Important to Safety .................................................................................. 2-2

2.4.2 Design Bases for SSCs Important to Safety ...................................................... 2-3

2.4.2.1 SNF Specifications ........................................................................................ 2-3

2.4.2.2 External Conditions ....................................................................................... 2-4

2.4.3 Design Criteria for Safety Protection Systems .................................................. 2-4

2.4.3.1 General .......................................................................................................... 2-4

2.4.3.2 Structural ....................................................................................................... 2-4

2.4.3.3 Thermal ......................................................................................................... 2-5

2.4.3.4 Shielding/Confinement/Radiation Protection ................................................. 2-5

2.4.3.5 Criticality ........................................................................................................ 2-5

2.4.3.6 Material Selection .......................................................................................... 2-5

2.4.3.7 Operating Procedures ................................................................................... 2-5

2.4.3.8 Acceptance Tests and Maintenance ............................................................. 2-5

2.4.3.9 Decommissioning .......................................................................................... 2-6
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2.5 Review Procedures ............................................................................................ 2-6

2.5.1 SSCs Important to Safety (MEDIUM Priority) .................................................... 2-6

2.5.2 Design Bases for SSCs Important to Safety ...................................................... 2-6

2.5.2.1 SNF Specifications (MEDIUM Priority) .......................................................... 2-8

2.5.2.2 External Conditions (MEDIUM Priority up to Natural Phenomena Events) ... 2-9

2.5.3 Design Criteria for Safety Protection Systems (MEDIUM Priority) .................. 2-14

2.6 Evaluation Findings .......................................................................................... 2-18

3 STRUCTURAL EVALUATION .............................................................................................. 3-1

3.1 Review Objective ................................................................................................ 3-1

3.2 Areas of Review ................................................................................................. 3-1

3.3 Regulatory Requirements ................................................................................... 3-2

3.4 Acceptance Criteria ............................................................................................ 3-2

3.4.1 Confinement Cask and Metallic Internals .......................................................... 3-3

3.4.1.1 Steel Confinement Cask ................................................................................ 3-3

3.4.1.2 Steel-Lined Concrete Confinement Cask ...................................................... 3-3

3.4.2 Other Structural System Components and Structures Important to Safety ....... 3-4

3.4.2.1 Steel Structures ............................................................................................. 3-4

3.4.2.2 Reinforced Concrete Structures .................................................................... 3-4

3.4.3 Other Structural Components Subject to NRC Approval ................................... 3-4

3.5 Review Procedures (HIGH Priority) .................................................................... 3-5

3.5.1 Confinement Cask and Metallic Internals .......................................................... 3-9

3.5.1.1 Scope ............................................................................................................ 3-9

3.5.1.2 Structural Design Criteria and Design Features ............................................ 3-9

3.5.1.3 Materials Related to Structural Evaluation (HIGH Priority) .......................... 3-13

3.5.1.4 Structural Analysis ....................................................................................... 3-14

3.5.2 Other System Components and Structures Important to Safety ..................... 3-27

3.5.2.1 Scope .......................................................................................................... 3-27

3.5.2.2 Structural Design Criteria and Design Features .......................................... 3-27

3.5.2.3 Structural Analysis ....................................................................................... 3-32

3.5.3 Other Structural Components Subject to NRC Approval (MEDIUM Priority) ... 3-37

3.5.3.1 Scope .......................................................................................................... 3-37

3.5.3.2 Structural Design Criteria and Design Features .......................................... 3-38

3.5.3.3 Materials Related to Structural Evaluation .................................................. 3-39

3.5.3.4 Structural Analysis ....................................................................................... 3-39

3.6 Evaluation Findings .......................................................................................... 3-40

3.7 Designations and Descriptions of Loads .......................................................... 3-41

3.7.1 Load Combinations for Steel and Reinforced Concrete Non-Confinement 

Structures ......................................................................................................... 3-46

APPENDIX 3A - COMPUTATIONAL MODELING SOFTWARE .............................................. 3-50

4 THERMAL EVALUATION ..................................................................................................... 4-1

4.1 Review Objective ................................................................................................ 4-1

4.2 Areas of Review ................................................................................................. 4-1

4.3 Regulatory Requirements ................................................................................... 4-1

4.4 Acceptance Criteria ............................................................................................ 4-2

4.4.1 Decay Heat Removal System ........................................................................... 4-2

4.4.2 Material and Design Limits ................................................................................ 4-2

4.4.3 Thermal Loads and Environmental Conditions .................................................. 4-3

4.4.4 Analytical Methods, Models, and Calculations .................................................. 4-3

4.5 Review Procedures ............................................................................................ 4-6 
 vii 

4.5.1 Decay Heat Removal System (HIGH Priority) ................................................... 4-6

4.5.2 Material and Design Limits (Priority - as indicated) ........................................... 4-7

4.5.3 Thermal Loads and Environmental Conditions (Priority - as indicated) ............. 4-8

4.5.4 Analytical Methods, Models, and Calculations (MEDIUM Priority) .................... 4-8

4.5.4.1 Configuration (HIGH Priority) ......................................................................... 4-9

4.5.4.2 Material Properties (MEDIUM Priority) ........................................................ 4-11

4.5.4.3 Boundary Conditions (Priority - as indicated) .............................................. 4-12

4.5.4.4 Computer Codes (HIGH Priority) ................................................................. 4-12

4.5.4.5 Temperature Calculations (Priority – as indicated) ...................................... 4-13

4.5.4.6 Pressure Analysis (LOW Priority) ................................................................ 4-15

4.5.4.7 Confirmatory Analysis (HIGH Priority) ......................................................... 4-16

4.6 Evaluation Findings .......................................................................................... 4-17

5 CONFINEMENT EVALUATION ............................................................................................ 5-1

5.1 Review Objective ................................................................................................ 5-1

5.2 Areas of Review ................................................................................................. 5-1

5.3 Regulatory Requirements ................................................................................... 5-2

5.4 Acceptance Criteria ............................................................................................ 5-2

5.4.1 Confinement Design Characteristics ................................................................. 5-2

5.4.2 Confinement Monitoring Capability.................................................................... 5-3

5.4.3 Nuclides with Potential for Release ................................................................... 5-3

5.4.4 Confinement Analyses ...................................................................................... 5-3

5.4.5 Supplemental Information ................................................................................. 5-4

5.5 Review Procedures ............................................................................................ 5-4

5.5.1 Confinement Design Characteristics (MEDIUM Priority) ................................... 5-4

5.5.1.1 Design Criteria ............................................................................................... 5-4

5.5.1.2 Design Features ............................................................................................ 5-6

5.5.2 Confinement Monitoring Capability (LOW Priority) ............................................ 5-7

5.5.3 Nuclides with Potential for Release (LOW Priority) ........................................... 5-8

5.5.4 Confinement Analyses (MEDIUM Priority) ...................................................... 5-10

5.5.4.1 Normal Conditions ....................................................................................... 5-12

5.5.4.2 Off-Normal Conditions (Anticipated Occurrences) ...................................... 5-12

5.5.4.3 Design-Basis Accident Conditions (Including Natural Phenomenon Events) ..... 

 ...................................................................................................................... 5-12

5.5.5 Supplemental Information ............................................................................... 5-13

5.6 Evaluation Findings .......................................................................................... 5-13

6 SHIELDING EVALUATION ................................................................................................... 6-1

6.1 Objective ............................................................................................................ 6-1

6.2 Areas of Review ................................................................................................. 6-1

6.3 Regulatory Requirements ................................................................................... 6-2

6.4 Acceptance Criteria ............................................................................................ 6-2

6.4.1 Shielding Design Description ............................................................................ 6-3

6.4.1.1 Design Criteria ............................................................................................... 6-4

6.4.1.2 Design Features ............................................................................................ 6-4

6.4.2 Radiation Source Definition ............................................................................... 6-4

6.4.2.1 Gamma Sources ............................................................................................ 6-4

6.4.2.2 Neutron Sources ............................................................................................ 6-5

6.4.3 Shielding Model Specification ........................................................................... 6-5

6.4.3.1 Configuration of Shielding and Source .......................................................... 6-5

6.4.3.2 Material Properties ........................................................................................ 6-5

6.4.4 Shielding Analyses ............................................................................................ 6-5
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6.4.4.1 Computer Codes ........................................................................................... 6-5

6.4.4.2 Flux-to-Dose-Rate Conversion ...................................................................... 6-6

6.4.4.3 Dose Rates .................................................................................................... 6-6

6.5 Review Procedures ............................................................................................ 6-7

6.5.1 Shielding Design Description ............................................................................ 6-7

6.5.1.1 Design Criteria (MEDIUM Priority) ................................................................. 6-7

6.5.1.2 Design Features (HIGH Priority) .................................................................... 6-9

6.5.2 Radiation Source Definition (HIGH Priority) ...................................................... 6-9

6.5.2.1 Initial Enrichment ........................................................................................... 6-9

6.5.2.2 Computer Codes for Radiation Source Definition ........................................ 6-10

6.5.2.3 Gamma Source ........................................................................................... 6-10

6.5.2.4 Neutron Source ........................................................................................... 6-11

6.5.2.5 Other Parameters Affecting the Source Term ............................................. 6-12

6.5.3 Shielding Model Specification (HIGH Priority) ................................................. 6-12

6.5.3.1 Configuration of the Shielding and Source .................................................. 6-12

6.5.3.2 Material Properties ...................................................................................... 6-13

6.5.4 Shielding Analyses .......................................................................................... 6-13

6.5.4.1 Computer Codes (MEDIUM Priority) ........................................................... 6-13

6.5.4.2 Flux-to-Dose-Rate Conversion (MEDIUM Priority) ...................................... 6-16

6.5.4.3 Dose Rates (MEDIUM Priority) .................................................................... 6-16

6.5.4.4 Confirmatory Calculations (HIGH Priority) ................................................... 6-18

6.5.5 Supplemental Information ............................................................................... 6-19

6.6 Evaluation Findings .......................................................................................... 6-19

7 CRITICALITY EVALUATION ................................................................................................ 7-1

7.1 Review Objective ................................................................................................ 7-1

7.2 Areas of Review ................................................................................................. 7-1

7.3 Regulatory Requirements ................................................................................... 7-1

7.4 Acceptance Criteria ............................................................................................ 7-2

7.5 Review Procedures ............................................................................................ 7-3

7.5.1 Criticality Design Criteria and Features (HIGH Priority) .................................... 7-3

7.5.2 Fuel Specification (HIGH Priority)...................................................................... 7-5

7.5.2.1 Non-Fuel Hardware ....................................................................................... 7-6

7.5.2.2 Fuel Condition ............................................................................................... 7-6

7.5.3 Model Specification (HIGH Priority) ................................................................... 7-7

7.5.3.1 Configuration ................................................................................................. 7-7

7.5.3.2 Material Properties ........................................................................................ 7-8

7.5.4 Criticality Analysis (Priority as indicated) ........................................................... 7-9

7.5.4.1 Computer Codes ........................................................................................... 7-9

7.5.4.2 Multiplication Factor ..................................................................................... 7-10

7.5.4.3 Benchmark Comparisons (HIGH Priority) .................................................... 7-11

7.5.5 Burnup Credit (HIGH Priority) .......................................................................... 7-12

7.5.6 Supplemental Information ............................................................................... 7-12

7.6 Evaluation Findings .......................................................................................... 7-13

8 MATERIALS EVALUATION .................................................................................................. 8-1

8.1 Review Objective ................................................................................................ 8-1

8.2 Areas of Review ................................................................................................. 8-2

8.3 Regulatory Requirements ................................................................................... 8-3

8.4 Review Procedures and Acceptance Criteria ..................................................... 8-3

8.4.1 General Review Considerations (HIGH Priority) ............................................... 8-3

8.4.2 Codes and Standards (HIGH Priority) ............................................................... 8-6
 ix 

8.4.2.1 Usage and Endorsement ............................................................................ 8-6

8.4.2.2 Code Case Use/Acceptability ..................................................................... 8-6

8.4.3 Environment (Priority – as indicated) ................................................................. 8-6

8.4.4 Drawings (MEDIUM Priority) ............................................................................. 8-7

8.4.5 Material Properties (MEDIUM Priority) .............................................................. 8-7

8.4.5.1 Structural Properties ...................................................................................... 8-7

8.4.5.2 Thermal Materials .......................................................................................... 8-8

8.4.6 Coastal Marine ISFSI Sites–Material Selections (MEDIUM Priority) ................. 8-8

8.4.7 Weld Design/Inspection (MEDIUM Priority) ....................................................... 8-9

8.4.7.1 Welding Codes–Background Discussion .................................................... 8-9

8.4.7.2 Weld Design and Testing ......................................................................... 8-10

8.4.7.3 Lid Welds and Closure Welds .................................................................. 8-11

8.4.7.4 Austenitic Stainless and Nickel-Base Alloy Steels Cask Design .............. 8-13

8.4.8 Galvanic/Corrosive Reactions (LOW Priority) ................................................. 8-13

8.4.8.1 Environmental considerations .................................................................. 8-13

8.4.8.2 Canister Contents ........................................................................................ 8-13

8.4.9 Creep Behavior of Aluminum Components (HIGH Priority) ............................. 8-14

8.4.10 Bolt Applications (MEDIUM Priority) ................................................................ 8-14

8.4.11 Exterior Protective Coatings (LOW Priority) .................................................... 8-14

8.4.11.1 Review Guidance ..................................................................................... 8-15

8.4.11.2 Scope of Coating Application ................................................................... 8-15

8.4.11.3 Coating Selection ..................................................................................... 8-15

8.4.11.4 Surface Preparation .................................................................................. 8-16

8.4.11.5 Coating Repairs ........................................................................................ 8-16

8.4.11.6 Coating Qualification Testing .................................................................... 8-17

8.4.12 Neutron Shielding (MEDIUM Priority) .............................................................. 8-18

8.4.12.1 Neutron Shielding Materials ..................................................................... 8-18

8.4.12.2 Assessing Previously Unreviewed (New) Neutron Shielding Materials .... 8-18

8.4.13 Criticality Control (HIGH Priority) ..................................................................... 8-19

8.4.13.1 Neutron-Absorbing/Poison Materials ........................................................ 8-20

8.4.13.2 Computation of Percent Credit for Boron-Based Neutron Absorbers ....... 8-20

8.4.13.3 Qualifying the Neutron Absorber Material Fabrication Process ................ 8-22

8.4.14 Concrete and Reinforcing Steel (LOW Priority) ............................................... 8-24

8.4.14.1 Embedment Materials ............................................................................... 8-24

8.4.14.2 Concrete Temperature Limits ................................................................... 8-25

8.4.14.3 Omission of Reinforcement ...................................................................... 8-25

8.4.15 Seals ............................................................................................................... 8-25

8.4.15.1 Metallic Seals (MEDIUM Priority) ............................................................. 8-26

8.4.15.2 Elastomeric Seals (LOW Priority) ............................................................. 8-26

8.4.16 Low Temperature Ductility and Fracture Control of Ferritic Steels 

(MEDIUM Priority) ............................................................................................ 8-26

8.4.17 Cladding .......................................................................................................... 8-27

8.4.17.1 Cladding Temperature Limits (MEDIUM Priority) ..................................... 8-27

8.4.17.2 Fuel Classification (HIGH Priority) ............................................................ 8-29

8.4.17.3 Reflood Analysis (HIGH Priority) .............................................................. 8-32

8.4.18 Prevention of Oxidation Damage During Loading of Fuel (MEDIUM Priority) . 8-32

8.4.19 Flammable Gas Generation (MEDIUM Priority) ............................................... 8-33

8.4.20 Helium Leakage Testing (MEDIUM Priority) ..................................................... 8-33

8.4.21 Periodic Inspections (LOW Priority)................................................................. 8-35

8.5 Evaluation Findings .......................................................................................... 8-36

8.6 Supplemental Information for Methods for Classifying Fuel (HIGH Priority) .... 8-37
 x 

8.7 Supplemental Information for Potential Rod Splitting Due to Exposure to an 

Oxidizing Atmosphere During Short-Term Cask Loading Operations in LWR or 

Other Uranium Oxide Based Fuel (MEDIUM Priority) ...................................... 8-41

8.7.1 Fuel Oxidation and Cladding Splitting ............................................................. 8-41

8.7.2 Data Base ........................................................................................................ 8-41

8.8 Supplemental Information for Background justification for Cladding Temperature 

Considerations for the Storage of Spent Fuel (MEDIUM Priority) .................... 8-43

8.8.1 Basis for Guidance .......................................................................................... 8-43

8.8.2 Review Guidance ............................................................................................ 8-45

8.9 Supplemental Information for the Design and Testing of Lid Welds on Austenitic 

Stainless Steel Canisters as Confinement Boundary for Spent Fuel Storage 

(MEDIUM Priority) ............................................................................................ 8-46

8.9.1 Basis for the Review ........................................................................................ 8-46

8.9.2 Helium Leak Test ............................................................................................ 8-47

9 OPERATING PROCEDURES EVALUATION ....................................................................... 9-1

9.1 Review Objective ................................................................................................ 9-1

9.2 Areas of Review ................................................................................................. 9-1

9.3 Regulatory Requirements ................................................................................... 9-2

9.4 Acceptance Criteria ............................................................................................ 9-2

9.4.1 Cask Loading .................................................................................................... 9-3

9.4.2 Cask Handling and Storage Operations ............................................................ 9-3

9.4.3 Cask Unloading ................................................................................................. 9-4

9.5 Review Procedures ............................................................................................ 9-4

9.5.1 Cask Loading (Priority - as indicated) ................................................................ 9-6

9.5.2 Cask Handling and Storage Operations (LOW Priority) .................................. 9-11

9.5.3 Cask Unloading (Priority – as indicated) ......................................................... 9-12

9.6 Evaluation Findings .......................................................................................... 9-14

10 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM EVALUATION ...................... 10-1

10.1 Review Objective .............................................................................................. 10-1

10.2 Areas of Review ............................................................................................... 10-1

10.3 Regulatory Requirements ................................................................................. 10-1

10.4 Acceptance Criteria .......................................................................................... 10-2

10.5 Review Procedures ......................................................................................... 10-2

10.5.1 Acceptance Tests (Priority – as indicated) ...................................................... 10-5

10.5.1.1 Structural/Pressure Tests ......................................................................... 10-5

10.5.1.2 Leak Tests (LOW Priority) ........................................................................ 10-7

10.5.1.3 Visual and Nondestructive Examination Inspections ................................ 10-7

10.5.1.4 Shielding Tests (LOW Priority) ................................................................. 10-9

10.5.1.5 Neutron Absorber Tests (HIGH Priority) ................................................. 10-10

10.5.1.6 Thermal Tests (LOW Priority) ................................................................. 10-12

10.5.1.7 DSS Identification (LOW Priority) ........................................................... 10-12

10.5.2 Maintenance Program (LOW Priority) ........................................................... 10-12

10.5.2.1 Inspection ............................................................................................... 10-12

10.5.2.2 Tests ....................................................................................................... 10-12

10.5.2.3 Repair, Replacement, and Maintenance ................................................ 10-13

10.6 Evaluation Findings ........................................................................................ 10-13

11 RADIATION PROTECTION EVALUATION ...................................................................... 11-1

11.1 Review Objective .............................................................................................. 11-1

11.2 Areas of Review ............................................................................................... 11-1
 xi 

11.3 Regulatory Requirements ................................................................................. 11-2

11.4 Acceptance Criteria .......................................................................................... 11-3

11.4.1 Radiation Protection Design Criteria and Features ......................................... 11-3

11.4.2 Occupational Exposures ................................................................................. 11-3

11.4.3 Exposures at or Beyond the Controlled Area Boundary .................................. 11-3

11.4.4 ALARA ............................................................................................................. 11-4

11.5 Review Procedures .......................................................................................... 11-4

11.5.1 Radiation Protection Design Criteria and Features for the Transfer Cask and 

Storage Cask (MEDIUM Priority) ...................................................................... 11-4

11.5.2 Occupational Exposures (MEDIUM Priority) ................................................... 11-6

11.5.3 Exposures at or Beyond the Controlled Area Boundary (MEDIUM Priority) .... 11-6

11.5.3.1 Normal Conditions .................................................................................... 11-7

11.5.3.2 Accident Conditions and Natural Phenomenon Events ............................ 11-8

11.5.4 ALARA (MEDIUM Priority)............................................................................... 11-8

11.5.4.1 Design Considerations ............................................................................. 11-8

11.5.4.2 Procedures and Engineering Controls ...................................................... 11-8

11.6 Evaluation Findings .......................................................................................... 11-9

12 ACCIDENT ANALYSES EVALUATION ............................................................................ 12-1

12.1 Review Objective .............................................................................................. 12-1

12.2 Areas of Review ............................................................................................... 12-1

12.3 Regulatory Requirements ................................................................................. 12-1

12.4 Acceptance Criteria .......................................................................................... 12-2

12.4.1 Dose Limits for Off-Normal Events .................................................................. 12-2

12.4.2 Dose Limit for Design-Basis Accidents............................................................ 12-3

12.4.3 Criticality .......................................................................................................... 12-3

12.4.4 Confinement .................................................................................................... 12-3

12.4.5 Recovery and Retrievability ............................................................................. 12-3

12.4.6 Instrumentation ................................................................................................ 12-3

12.5 Review Procedures .......................................................................................... 12-4

12.5.1 Cause of the Event (MEDIUM Priority) ............................................................ 12-5

12.5.2 Detection of the Event (MEDIUM Priority) ....................................................... 12-5

12.5.3 Summary of Event Consequences and Regulatory Compliance 

(MEDIUM PRIORITY) ...................................................................................... 12-5

12.5.4 Corrective Course of Action (MEDIUM Priority) .............................................. 12-5

12.6 Evaluation Findings .......................................................................................... 12-6

13 TECHNICAL SPECIFICATIONS AND OPERATING CONTROLS AND LIMITS 

EVALUATION ................................................................................................................... 13-1

13.1 Review Objective .............................................................................................. 13-1

13.2 Areas of Review ............................................................................................... 13-1

13.3 Regulatory Requirements ................................................................................. 13-2

13.4 Acceptance Criteria .......................................................................................... 13-2

13.4.1 Functional/Operating Limits, Monitoring Instruments, and Limiting Control 

Settings ............................................................................................................ 13-3

13.4.2 Limiting Conditions .......................................................................................... 13-3

13.4.3 Surveillance Requirements ............................................................................. 13-3

13.4.4 Design Features .............................................................................................. 13-3

13.4.5 Administrative Control ..................................................................................... 13-5

13.5 Review Procedures (HIGH Priority) .................................................................. 13-5

13.6 Evaluation Findings .......................................................................................... 13-8
 xii 

14 QUALITY ASSURANCE EVALUATION ........................................................................... 14-1

14.1 Review Objective .............................................................................................. 14-1

14.2 Areas of Review ............................................................................................... 14-1

14.3 Regulatory Requirements ................................................................................. 14-1

14.4 Acceptance Criteria .......................................................................................... 14-1

14.5 Review Procedures (All items in this section are HIGH Priority) ...................... 14-1

14.5.1 Quality Assurance Organization ...................................................................... 14-3

14.5.2 Quality Assurance Program ............................................................................ 14-4

14.5.3 Design Control ................................................................................................. 14-6

14.5.4 Procurement Document Control ...................................................................... 14-7

14.5.5 Instructions, Procedures, and Drawings .......................................................... 14-8

14.5.6 Document Control ........................................................................................... 14-8

14.5.7 Control of Purchased Material, Equipment, and Services ............................... 14-9

14.5.8 Identification and Control of Materials, Parts, and Components ................... 14-11

14.5.9 Control of Special Processes ........................................................................ 14-11

14.5.10 Licensee Inspection ........................................................................................ 14-12

14.5.11 Test Control .................................................................................................... 14-13

14.5.12 Control of Measuring and Test Equipment ..................................................... 14-13

14.5.13 Handling, Storage, and Shipping Control ....................................................... 14-14

14.5.14 Inspection, Test, and Operating Status .......................................................... 14-14

14.5.15 Nonconforming Materials, Parts, or Components........................................... 14-15

14.5.16 Corrective Action ............................................................................................ 14-15

14.5.17 Quality Assurance Records ............................................................................ 14-16

14.5.18 Audits ............................................................................................................. 14-17

14.6 Evaluation Findings ........................................................................................ 14-18

APPENDIX A CONSOLIDATED REFERENCES ................................................................... A-1

APPENDIX B PROCESS FOR PRIORITIZING THE STANDARD REVIEW PLAN FOR DRY 

STORAGE SYSTEMS ....................................................................................................... B-1

APPENDIX C LIST OF ISGs 1 TO 22 WITH THOSE INCORPORATED INTO NUREG-1536 

IDENTIFIED ....................................................................................................................... C-1

APPENDIX D PUBLIC COMMENTS RECEIVED AND THEIR DISPOSITION.........D1

LIST OF FIGURES 

Figure 1-1 Overview of Safety Evaluation ................................................................................. 1-5

Figure 2-1 Overview of Principal Design Criteria Evaluation .................................................... 2-7

Figure 3-1 Overview of the Structural Evaluation ..................................................................... 3-7

Figure 4-1 Overview of the Thermal Evaluation ........................................................................ 4-5

Figure 5-1 Overview of the Confinement Evaluation ................................................................ 5-5

Figure 6-1 Overview of the Shielding Evaluation ...................................................................... 6-8

Figure 7-1 Overview of Criticality Evaluation ............................................................................. 7-4

Figure 8-1 Overview of Materials Evaluation ............................................................................ 8-5

Figure 8-2 Relationship of Spent Fuel Populations ................................................................ 8-39

Figure 9-1 Overview of Operating Procedures Evaluation ........................................................ 9-5

Figure 10-1 Overview of Acceptance Test Review Evaluation ............................................... 10-3

Figure 10-2 Overview of Maintenance Program Review Evaluation ........................................ 10-4

Figure 11-1 Overview of the Radiation Protection Evaluation ................................................. 11-5

Figure 12-1 Overview of Accident Analysis Evaluation ........................................................... 12-4
 xiii 

Figure 13-1 Provision Example ................................................................................................ 13-4

Figure 13-2 Overview of Technical Specifications and Operating Controls Evaluation .......... 13-6

Figure 14-1 Quality Assurance Evaluation ............................................................................. 14-3

LIST OF TABLES 

Table 1-1 Relationship of Regulations and Areas of Review .................................................... 1-2

Table 2-1 Relationship of 10 CFR Part 72 Regulations and Areas of Review .......................... 2-2

Table 2-2 Outline of Design Criteria and Bases for DSS ........................................................ 2-15

Table 3-1 Relationship of Regulations and Areas of Review .................................................... 3-2

Table 3-2 Loads and Their Descriptions ................................................................................. 3-42

Table 3-3 Load Combinations for Steel and Reinforced Concrete Non-Confinement Structures

 ................................................................................................................................................. 3-46

Table 4-1 Relationship of Regulations and Areas of Review .................................................... 4-2

Table 5-1 Relationship of Regulations and Areas of Review .................................................... 5-2

Table 5-2 Fractions of Radioactive Materials Available for Release from Spent Fuel .............. 5-9

Table 6-1 Relationship of Regulations and Areas of Review .................................................... 6-2

Table 7-1 Relationship of Regulations and Areas of Review .................................................... 7-2

Table 8-1 Relationship of 10 CFR Part 72 Regulations and Areas of Review .......................... 8-3

Table 9-1 Relationship of Regulations and Areas of Review .................................................... 9-2

Table 10-1 Relationship of Regulations and Areas of Review ................................................ 10-2

Table 11-1 Relationship of 10 CFR Part 20 Regulations and Areas of Review ...................... 11-2

Table 11-2 Relationship of 10 CFR Part 72 Regulations and Areas of Review ...................... 11-2

Table 12-1 Relationship of Regulations and Areas of Review ................................................ 12-2

Table 13-1 Relationship of Regulations and Areas of Review ................................................ 13-2

 xv 

ACRONYMS AND ABBREVIATIONS 

 

ACI American Concrete Institute 

ADE annual dose equivalent 

AISC American Institute of Steel Construction 

ALARA as low as is reasonably achievable 

ANL Argonne National Laboratory 

ANS American Nuclear Society 

ASCE American Society of Civil Engineers 

ANSI American National Standards Institute 

API American Petroleum Institute 

APSR axial power shaping rod 

ASD allowable stress design 

ASME American Society of Mechanical Engineers 

ASTM American Society for Testing and Materials 

AWWA American Water Works Association 

AWS American Welding Society 

 

B&PV boiler and pressure vessel 

BPRA burnable poison rod assembly 

BR breathing rate 

BWR boiling-water reactor 

 

CDE committed dose equivalent 

CEA control element assembly 

CEDE committed effective dose equivalent 

CFD computational fluid dynamics 

CFR U.S. Code of Federal Regulations 

CoC Certificate of Compliance 

CSFM Commercial Spent Fuel Management Program 

 

DBA design-basis accident 

DBE design-basis event 

DCF dose conversion factor 

DSS dry storage system 

DDE deep dose equivalent 

DE design earthquake 
 xvi 

DLF design load factor 

DOE U.S. Department of Energy 

 

EPA Environmental Protection Agency 

EPRI Electric Power Research Institute 

 

FR Federal Registry 

 

g gram 

Gr Grashof 

GTCC greater than Class C 

Gy Gray 

Gz Graetz 

 

HAC hypothetical accident condition 

HAZ heat affected zone 

HTGR high-temperature gas-cooled reactor 

H/U hydrogen-to-uranium 

 

IBC International Building Code 

ICBO International Conference of Building Officials 

ICC International Code Council 

ICRP International Commission on Radiological Protection 

INEL Idaho National Engineering Laboratory 

ISFSI Independent Spent Fuel Storage Installation 

ISG Interim Staff Guidance 

 

LANL Los Alamos National Laboratory 

LCO limiting condition of operations 

LDE lens dose equivalent 

LLNL Lawrence Livermore National Laboratory 

LRFD load resistance factor design 

LT leak testing 

LWR light water reactor 

 

mJ milliJoule 

mm millimeter 

MNOP maximum normal operating pressure 
 xvii 

MPa megapascal 

ms millisecond 

MT magnetic particle examination 

 

N Newton 

NDE nondestructive examination 

NDT nil-ductility transition 

NEI Nuclear Energy Institute 

NFPA National Fire Protection Association 

NOAA National Oceanic and Atmospheric Administration 

NRC United States Nuclear Regulatory Commission 

NRPB National Radiation Protection Board 

NRR Office of Nuclear Reactor Regulation 

 

OBE operating-basis earthquake 

OFA optimized fuel assembly 

ORNL Oak Ridge National Laboratory 

 

PNL Pacific Northwest Laboratory 

PT liquid (dye) penetrant examination 

PWHT preheat and post-weld heat treatment 

PWR pressurized-water reactor 

 

QA Quality Assurance 

QAPD Quality Assurance Program Description 

QC quality control 

 

RAI request for additional information 

RC reinforced concrete 

RCCA rod cluster control assembly 

RG Regulatory Guide 

RSICC Radiation Safety Information Computational Center 

RT radiographic examination 

 

SAR Safety Analysis Report 

SDE shallow (skin) dose equivalent 

SEM scanning electron microscopy 

SER Safety Evaluation Report 
 xviii 

SFST Division of Spent Fuel Storage and Transportation 

SI système international (d'unités) (International System of Units)

SNF spent nuclear fuel 

SNT American Society for Nondestructive Testing 

SRP Standard Review Plan 

SSC structures, systems, and components 

SSE safe shutdown earthquake 

Sv Sievert 

 

TEDE total effective dose equivalent 

TEM transmission electron microscopy 

TODE total organ dose equivalent 

TS Technical Specification 

TSAR Topical Safety Analysis Report 

 

UBC Uniform Building Code 

UK United Kingdom 

UT ultrasonic examination 

 

VT visual examination 

 
 xix 

UNITS 

Btu/hr.ft.°F British thermal units per hour-foot-degree Fahrenheit

°C degrees Centigrade 

Ci/cm3

 Curies per cubic centimeters 

Ci/s Curies per second 

cm3

/s cubic centimeters per second

°F degrees Fahrenheit 

ft feet 

ft/s feet per second 

ft3

 cubic feet 

ft3

/s cubic feet per second

g/cm3 grams per cubic centimeters

GWd/MTU GigaWatt days per Metric Ton Uranium

in. inches 

K Kelvin 

kg kilogram 

kgf/cm2

 kilograms force per square centimeters

kPa kiloPascal 

ksi thousand pounds per square inch

kW kilowatts 

lb pounds

m meters 

m2

 square meters 

m3

 cubic meters 

m3

/s cubic meters per second

m/s meters per second 

mCi millicuries (one-thousandth of a curie)

MeV million electron volts

mg milligram (one-thousandth of a gram)

mm millimeters (one-thousandth of a meter)

MPa MegaPascal (million Pascals)

mrem millirem (one-thousandth of a rem)

mSv millisievert (one-thousandth of a sievert)

MWd/MTU MegaWatt days per Metric Ton Uranium

pCi/m3 picocurie (one-trillionth of a curie)/cubic meter

PM10 particulate matter (less than 10 microns in diameter)

ppm parts per million 
 xx 

psi pounds per square inch

s second 

Sv sievert 

μCi microcurie (one-millionth of a curie)

μCi/cm2 microcurie per square centimeter

W/m.K Watts per meter - Kelvin

 
 xxi 

GLOSSARY 

The following terms are defined here by the staff for the purpose of this document. 

Acceptance Test. Tests conducted by the applicant to ensure that material or component 

produced in a given production run is in compliance with the material or design requirements of 

the application. Acceptance tests are also used to ensure that the process is operating in a 

satisfactory manner by using statistical data for selected measurable parameters. 

Accident-Level. A term used to include both design-basis accidents and design-basis natural 

phenomenon events and conditions. 

Areal Density. Mass per unit area, usually expressed in grams per square centimeters (g/cm2

). 

In this document, this term is used to describe the distribution of neutron absorber content in a 

material. 

Adequate Margin. In the design of structures, systems, and components, the margin for safety 

is achieved by satisfying the acceptance criteria of the codes and standards for the specified 

design criteria loads, and the design basis (performance requirements). The reviewer must 

judge if the calculated design bases values require any margins with respect to the acceptance 

criteria of the codes and standards. This may depend on the uncertainties associated with the 

calculation of predicted design bases values (stress, displacements, etc.) used as reference for 

the performance of the structures. 

As Low As is Reasonably Achievable (ALARA). Making every reasonable effort to maintain 

exposures to radiation as far below the dose limits in 10 CFR Part 20 as is practical and 

consistent with the purpose for which the licensed activity is undertaken taking into account the 

state of technology, the economics of improvements in relation to state of technology, the 

economics of improvements in relation to benefits to the public health and safety, other societal 

and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed 

materials in the public interest (10 CFR 20.1003). Per 10 CFR 72.3, ALARA means as low as 

reasonably achievable taking into account the state of technology, and the economics of 

improvement in relation to: (1) benefits to the public health and safety, (2) other societal and 

socioeconomic considerations, and (3) the utilization of atomic energy in the public interest. 

Benchmarking. Establishment of the bias of a computer code for a particular application by 

comparison of the calculated results with the measured results of relevant representative 

experiments. For purposes of criticality analyses, benchmarking is the process of establishing 

the bias of the calculational method, which includes aspects such as the computer code, cross 

sections set, analyst’s technique, and analysis assumptions. 

Bias. ANSI/ANS-8.1 defines bias as “a measure of the systematic differences between 

calculational method results and experimental data” and uncertainty in the bias as “a measure 

of both the accuracy and the precision of the calculations and the uncertainty of the 

experimental data.” See NUREG/CR-6361 for further discussion of bias. Bias defined as the 

average of the differences between results and measurements may be acceptable, provided 

that one adequately considers the variation in the differences. 

Burnable Poison Rod Assembly (BPRA). An assembly of poison rods used to absorb neutrons 

created in the nuclear reactor to control the power produced in the associated fuel assembly 

during the early core life. The BPRs are inserted into the fuel assemblies through the upper end 
 xxii 

fittings of the assembly and held in place against lift forces in the core by a retainer mechanism. 

BPRs within the spent fuel assembly envelope may be approved for storage in a dry storage 

system as part of the spent fuel assembly. 

Burnup. The measure of the thermal power produced in a specific amount of nuclear fuel 

through fission, usually expressed in units of MWd/MTU (megawatt days per metric ton of 

uranium). For the purpose of assessing the allowable contents, the maximum burnup(s) of the 

fuel should be specified in terms of the average burnup of the entire fuel assembly (i.e. 

assembly average). For the purpose of assessing fuel cladding integrity in the materials review, 

the rod with the highest burnup within the fuel assembly should be specified in terms of peak 

rod average burnup. 

Calculational Method. The calculational procedures – mathematical equations, approximations, 

assumptions, and associated numerical parameters (e.g., cross sections) – that yield the 

calculated results (ANSI/ANS-8.1-1998). 

Canister. In a dry storage system for spent nuclear fuel, a metal cylinder that is sealed at both 

ends and may be used to perform the function of confinement. Typically, a separate overpack 

performs the radiological shielding and physical protection function. 

Canning. To store damaged or consolidated spent nuclear fuel or nuclear fuel debris in a 

separate container and confine it in such a way that degradation of the fuel during storage will 

not pose operational safety problems with respect to its removal from storage 

[10 CFR 72.122(h)(1)]. 

Cask. In a dry storage system using the cask design for spent nuclear fuel, a passive standalone component that performs the functions of confinement, radiological shielding, decay heat 

removal, and physical protection of spent fuel during normal, off-normal, and accident-level 

conditions (NUREG-1571). 

Certificate of Compliance. The certificate issued by the NRC that approves the design of a 

spent nuclear fuel storage cask in accordance with the provisions of Subpart L of 10 CFR 72 

(10 CFR 72.3). 

Code. A generic reference to a national or “consensus” code, standard, and specification, or 

specifically to the ASME Boiler and Pressure Vessel Code (ASME B&PV Code). 

Committed Dose Equivalent (HT,50). The dose equivalent to organs or tissues of reference (T) 

that will be received from an intake of radioactive material by an individual during the 50-year 

period following the intake (10 CFR 20.1003). 

Confinement. The ability to prevent the release of radioactive substances into the environment 

(NUREG-1571). 

Confinement System. Those systems, including ventilation, that act as barriers between areas 

containing radioactive substances and the environment (10 CFR 72.3). 

Confirmatory Calculations. Calculations made by the reviewer to determine whether the cask 

design and specifications meet the requirements of the Code of Federal Regulations. These 

calculations do not replace the design calculations and are not intended to endorse the 

applicant’s calculations. 
 xxiii 

Construction. Includes materials, design, fabrication, installation, examination, testing, 

inspection, and certification as required in the manufacture and installation of components. 

Control Element Assembly (CEA) – An assembly of neutron poison elements used to control the 

reactor power during operations, if needed, and to provide shutdown capability. This 

component is designed for operations within the fuel assembly envelope, and when stored with 

spent fuel, fits within that envelope. 

Controlled Area. For an independent spent fuel storage installation (ISFSI), that area 

immediately surrounding the ISFSI for which the licensee exercises authority over its use and 

within which ISFSI operations are performed (10 CFR 72.3). For a nuclear power plant, that 

area outside of a restricted area but inside the site boundary to which access can be limited by 

the licensee for any reason (10 CFR 20.1003). 

Criticality. A measurement of the state of a fission system. 

Curie. The basic unit of radioactivity. A curie is equal to 37 billion (3.7 X 1010) disintegrations 

per second. 

Damaged Fuel. Spent nuclear fuel is considered damaged for storage purposes if it cannot 

fulfill its regulatory or design function. Specific conditions that define damaged fuel are provided 

in Section 8.4.17.2 of this SRP. Section 8.6, Supplemental Information for Methods for 

Classifying Fuel, provides methods for classifying spent nuclear fuel as damaged. 

Damaged-Fuel Can. A metal enclosure that is sized to confine one damaged spent fuel 

assembly. A fuel can for damaged spent fuel with damaged spent-fuel assembly contents must 

satisfy fuel-specific and system-related functions for undamaged SNF required by the applicable 

regulations. 

Degradation. Any change in the properties of a material that adversely affects the behavior of 

that material; adverse alteration (ASTM C1174-97). 

Design Bases. The information that identifies the specific functions to be performed by a 

structure, system, or component (e.g., spent fuel storage cask) and the specific values or 

ranges of values chosen for controlling parameters as reference bounds for design. 

Design Earthquake. The design earthquake ground motion for a site where a cask system may 

be used that is determined in accordance with 72.102 or 72.103. 

Design Event (I, II, III, or IV). Conditions and events as defined and used for an independent 

spent fuel storage installation in ANSI/ANS 57.9. 

Double Contingency Principle. A design principle requiring that at least two unlikely, 

independent, and concurrent or sequential changes in conditions essential to nuclear criticality 

safety must occur before a criticality accident is possible (10 CFR 72.124(a)). 

Exclusion Area. At a nuclear reactor site, the area surrounding the reactor in which the reactor 

licensee has the authority to determine all activities including exclusion or removal of personnel 

and property from the area. This area may be traversed by a highway, railroad, or waterway 

provided these are not so close to the facility as to interfere with normal operations of the 
 xxiv 

facility, and provided appropriate and effective arrangements are made to control traffic on the 

highway, railroad, or waterway, in case of emergency, to protect the public health and safety. 

Residence within the exclusion area shall normally be prohibited. In any event, residents shall 

be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor 

may be permitted in an exclusion area under appropriate limitations, provided that no significant 

hazards to the public health and safety will result (10 CFR 50.2). 

Gray (Gy). The SI unit of absorbed dose. 1 Gy is equal to 100 rad. 

Hard Receiving Surface. For a horizontal or vertical drop, need not be an unyielding surface; 

rather, the receiving surface may be modeled as a reinforced concrete pad on engineered fill. 

High Burnup Fuel. Spent nuclear fuel with burnups (see “Burnup”) generally exceeding 

45 GWd/MTU. 

Hoop Stress. The tensile stress in the cladding wall in the circumferential orientation. 

Important Confinement Features. See “important to safety.” 

Important to Safety, “Important to Nuclear Safety,” or “Structures, Systems, and Components 

Important to Safety.” Those features of a dry storage system that have one or more of the 

following functions: (1) maintain the conditions required to store spent nuclear fuel safely; 

(2) prevent damage to the spent nuclear fuel cask during handling or storage; or (3) provide 

reasonable assurance that spent nuclear fuel can be received, handled, containerized, stored, 

and retrieved without undue risk to the health and safety of the public. ANSI/ANS 57.9 uses the 

term “important confinement features”; however, NRC does not find this term acceptable. Per 

Regulatory Guide 3.60, Design of an Independent Spent Fuel Storage Installation (Dry Storage), 

“important to safety” should be substituted for “important confinement features” in the standard. 

Interim Staff Guidance (ISG). Supplemental information that clarifies important aspects of 

regulatory requirements. An ISG provides NRC review guidance to NRC Staff in a timely 

manner until standard review plans are revised accordingly. 

Low Burnup Fuel. Spent nuclear fuel with burnups (see “Burnup”) generally less than 

45 GWd/MTU. 

Margin of Safety, or MofS. This term may be defined, through a factor of safety, f.s = 

capacity/demand, as MofS = F.S.(capacity/demand)-1 (with minimum acceptable MofS> 0.0).” 

Misloading. The placement in a cask of spent nuclear fuel in a configuration not supported by 

the cask’s design basis or technical specifications. Also, the placement in a cask of spent 

nuclear fuel with characteristics that do not meet the characteristics of the cask’s allowable 

contents. 

Monitoring. Testing and data collection to determine the status of a dry storage system and to 

verify the continued efficacy of the system on the basis of measurements of specified 

parameters including temperature, radiation, and functionality and/or characteristics of 

components of the system. With respect to radiation, per 10 CFR 20.1003, monitoring means 

the measurement of radiation levels, concentrations, surface area concentrations or quantities 

of radioactive material, and the use of the results of these measurements to evaluate potential 

exposures and doses. 
 xxv 

Neutron Absorber. Also known as “poison.” Materials that have high neutron absorption cross 

section and are used to absorb neutrons to make a fission system less reactive. They are used 

to ensure subcriticality during normal/offnormal/accident-level conditions in containers of fissile 

materials. 

Nondestructive Examination (NDE). Testing, examination, and/or inspection of a component 

that does not affect the functionality and performance of the component. NDE can be broadly 

divided into three categories: visual, surface, and volumetric examinations. Additional 

information may be found in the ASME B&PV Code, Section V, Nondestructive Examination, 

Appendix A. 

NDE-related terms in order of increasing severity: 

 Discontinuity: An interruption in the normal physical structure of a material. 

Discontinuities may be unintentional (such as those formed inadvertently 

during the fabrication process) or intentional (such as a drilled hole). 

 Indication: Sign of a discontinuity observed when using an NDE method. 

 Flaw: An imperfection in an item or material which may or may not be harmful. 

 Defect: A flaw that, due to its size, shape, orientation, location, or other 

properties, is rejectable to the applicable construction code. Defects may 

be detrimental to the intended service of a component and the component 

must be repaired or replaced. 

Common NDE examination methods include: 

 LT leak testing 

 MT magnetic particle examination 

 PT liquid penetrant examination 

 RT radiographic examination 

 UT ultrasonic examination 

 VT visual examination 

Non-Fuel Hardware. Hardware that is not an integral part of a fuel assembly. Burnable Poison 

Rod Assembly (BPRA), Control Element Assembly (CEA), Thimble Plug Assembly (TPA), etc. 

are typical non-fuel hardware. 

Normal Events and Conditions. Conditions that are intended operations, planned events, and 

environmental conditions, that are known or reasonably expected to occur with high frequency 

during storage operations The maximum level of an event or condition it that expected to 

routinely occur. The cask system is expected to remain fully functional and to experience no 

temporary or permanent degradation from normal operations, events and conditions. Specific 

normal conditions to be addressed are evaluated for each dry storage system and are 

documented in a safety analysis report for that system. 

Normal Means. The ability to move a fuel assembly and its contents by the use of a crane and 

grapple used to move undamaged assemblies at the point of cask loading. The addition of 

special tooling or modifications to the assembly to make the assembly suitable for lifting by 
 xxvi 

crane and grapple does not preclude the assembly as being considered moveable by normal 

means. 

Off-Normal Events or Conditions. The maximum level of an event or condition that although not 

occurring regularly can be expected to occur with moderate frequency and for which there is a 

corresponding maximum specified resistance, limit of response, or requirement for a given level 

of continuing capability. “Off-Normal” events and conditions are similar to “Design Event II” of 

ANSI/ANS 57.9. An independent spent fuel storage installation structure, system, or component 

is expected to experience off-normal events and conditions without permanent deformation or 

degradation of capability to perform its full function (although operations may be suspended or 

curtailed during off-normal conditions) over the full license period. 

Preferential Loading. A non-uniform loading configuration of spent fuel assemblies within a dry 

storage system, that is typically specified by assigning a fuel zone designation to each basket 

cell, and specifying limiting nuclear and physical parameters of SNF assemblies that can be 

loaded into each zone. Preferential loading is often used as a means to optimize allowable SNF 

parameters (e.g. burnup, cooling time, decay heat), while satisfying the shielding, criticality, and 

thermal performance objectives of the cask system. 

Qualification Test. A test, or series of tests, that is conducted at least once for a given 

manufacturing process and set of material specifications to demonstrate the quality and 

durability of the component such as neutron absorber product over its licensed service life. 

Rad. The unit of absorbed dose. 1 rad is equal to the absorption of 100 ergs per gram. 

Ready Retrieval. The ability to move a canister containing spent fuel to either a transportation 

package or to a location where the spent fuel can be removed. Ready retrieval also means 

maintaining the ability to handle individual or canned spent fuel assemblies by the use of normal 

means 

Real Individual. A person who is not a nuclear worker and who is at or beyond the controlled 

area of an independent spent fuel storage installation, a nuclear power plant, or other nuclear 

facility. For example, a real individual may be anyone living, working, or recreating close to the 

facility for a significant portion of the year. 

Reasonable Assurance. NRC staff base their decisions on the adequacy of a dry storage 

system design to protect public health and safety on a variety of factors including: technical 

evaluations, test and operational data, compliance with NRC requirements, and insights from 

operational safety events. 

Recovery. The capability to return the stored radioactive material to a safe condition after an 

accident event without endangering public health and safety. This generally means ensuring 

that any potential release of radioactive materials to the environment or radiation exposures is 

not in excess of the limits in 10 CFR Part 20 during post-accident recovery operations. 

 

Rem. The special unit of any of the quantities expressed as dose equivalent. The dose 

equivalent in rems is equal to the absorbed dose in rads multiplied by the quality factor 

(1 rem = 0.01 sievert) (10 CFR 20.1004). 
 xxvii 

Restricted Area. An area to which access is limited by the licensee for the purpose of protecting 

individuals against undue risks from exposure to radiation and radioactive materials. Restricted 

area does not include areas used as residential quarters, but separate rooms in a residential 

building may be set apart as a restricted area (10 CFR 20.1003). 

Retrievability. In accordance with 10 CFR 72.122(l), storage systems must be designed to allow 

ready retrieval of spent fuel, high-level radioactive waste, and reactor-related GTCC waste for 

further processing or disposal._ 

Safety Analysis Report (SAR). In the context of this standard review plan, the report submitted 

to the NRC staff by a certificate applicant to present information related to the design of a dry 

storage system. This document provides the justification and analyses to demonstrate that the 

design meets the requirements and acceptance criteria. 

Safety Evaluation Report (SER). In the context of this standard review plan, the report prepared 

by the NRC staff to present findings and recommendations relating to the acceptability of an 

applicant’s safety analysis and other required documents submitted as part of a certificate 

application. The SER also identifies the bases for those recommendations and the 

recommended technical specifications (“operating controls and limits” or “conditions of use”). 

Safety Functions. The functions that dry storage system structures, systems, and components 

important to safety are designed to maintain include: 

 • Protection against environmental conditions, 

 • Content Temperature Control, 

 • Radiation Shielding, 

 • Confinement, 

 • Sub-criticality control, 

 • Retrievability. 

Sievert (Sv). The SI unit of any of the quantities expressed as dose equivalent. 1 Sv equals 

100 rem. The dose equivalent in sieverts equals the absorbed dose in grays multiplied by the 

quality factor (10 CFR 20.1004). 

Spent Nuclear Fuel, (SNF). Nuclear fuel that has been withdrawn from a nuclear reactor 

following irradiation, has undergone at least one year’s decay since being used as a source of 

energy in a power reactor, and has not been chemically separated into its constituent elements 

by reprocessing. Spent fuel includes the special nuclear material, byproduct material, source 

material, and other radioactive materials associated with fuel assemblies (10 CFR 72.3). 

Subcritical. The state at which the number of fission neutrons decreases with time and the 

effective neutron multiplication factor (keff) is less than unity. 

 

Supplemental Shielding. At an independent spent fuel storage installation, an engineered 

radiation shield (principally neutron and gamma radiation) such as an earthen berm or concrete 

wall. Supplemental shielding shall be deemed as component(s) important to safety and be 

specified in the Technical Specifications as a condition for use of the system as designed, if 

credited in the shielding analyses for meeting 72.104(a) or 72.106(b) requirements. 

Thimble Plug Assembly (TPA) – An assembly of short rods used to restrict the flow of coolant 

through a fuel assembly by being inserted into the assembly’s guide tubes. This component is 
 xxviii 

designed for operations within the fuel assembly envelope, and when stored with spent fuel, fits 

within that envelope. 

Total Effective Dose Equivalent (TEDE). The sum of the deep-dose equivalent for external 

exposures and the committed effective dose equivalent for internal exposures 

(10 CFR 20.1003). 

Unrestricted Area. An area to which access is neither limited nor controlled by the licensee 

(10 CFR 20.1003). 

Validation. Demonstration of the validity of a computer code for use in a general area of 

application by comparison of the code’s calculational results with the measured results from a 

variety of experiments spanning the area of intended applications. 

Volume Percent. The percent of a mole of the material that is present in a volume equal to the 

standard volume for the material as a gas; the volume occupied by one mole of the material as 

a gas at standard conditions for gases (760 mm Hg [760 torr] pressure and 0°C [32°F] 

temperature). 
 1 

INTRODUCTION 

This document is a Standard Review Plan (SRP). It is intended to provide guidance to the NRC 

staff conducting the safety review of an application for a spent fuel dry storage system (DSS) for 

facilities storing spent fuel under the general license authorized by 10 CFR 72.210. A general 

license authorizes a nuclear power plant licensee to store spent nuclear fuel (SNF) in NRCapproved casks at a site that is licensed to operate a power reactor under 10 CFR Part 50. 

This SRP was developed to promote a consistent regulatory review of an application for a DSS, 

present a basis for the review scope, and identify acceptable approaches to meeting regulatory 

requirements. 

This introduction provides an overview of the DSS and the Safety Analysis Report (SAR) review 

process, and assists the project manager in the coordination of the review effort. It is also 

designed to help individual technical reviewers understand how their specific review should be 

coordinated and integrated with other disciplines to produce a complete Safety Evaluation 

Report (SER). 

This SRP may be revised and updated as the need arises to clarify the content, correct errors, 

or incorporate modifications approved by the Director of the Division of Spent Fuel Storage and 

Transportation (SFST). Comments, suggestions for improvement, and notices of errors or 

omissions will be considered by and should be sent to the Director, Division of Spent Fuel 

Storage and Transportation, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear 

Regulatory Commission, Washington, DC 20555-0001. 

Use of Dry Storage Systems

In accordance with the requirements set forth in 10 CFR 72.212, a DSS may be used to store 

SNF in an independent spent fuel storage installation (ISFSI) under a general license. At 

present, any holder of an active reactor operating license under Title 10, Part 50, of the U.S. 

Code of Federal Regulations (10 CFR Part 50) has the authority to construct and operate an 

ISFSI using NRC-approved cask designs under the provisions of the general license. 

The DSS safety review is primarily based on the information provided by an applicant, or cask 

vendor, in a SAR. Section 72.230 of 10 CFR Part 72 requires inclusion of a SAR in each 

application for approval of SNF cask storage design. Before submitting a SAR, an applicant 

should have designed the DSS considering as-low-as-is-reasonably-achievable (ALARA) 

principles for radiation protection and analyzed it in sufficient detail to conclude that it can be 

properly fabricated and safely operated without endangering the health and safety of the public. 

The SAR is the principal document in which the applicant provides the information on the design 

and operational features and their associated technical bases. The reviewers need to 

understand the design and operational features and their technical bases, including but not 

limited to the selection of materials and geometries, mathematical models and equations used, 

computer models and calculated results in order to be able to draw conclusions that the storage 

cask is acceptable for use. 

Technical Review Oversight 

Cask designers are responsible for the safety of the cask design, and the cask users are 

responsible for safely operating the cask system at Part 50 reactor sites and complying with 

appropriate safety regulations. The mission of the regulator is to license and regulate the use of 
 2 

each DSS and ensure adequate protection of public health and safety. The value of the NRC 

review team is its independent expertise in identifying and resolving potential design or 

operational deficiencies; potential analytical errors; significant uncertainties in novel design 

approaches; or other non-compliance problems. If otherwise left unchecked by the designer, 

user and regulator, these issues could potentially lead to the unsafe or non-compliant use of the 

DSS. 

Several considerations may influence the depth and rigor that is needed for a reasonable 

assurance determination of both safety and compliance. These include the novelty of the 

design (as compared to existing designs); safety margins; operational experience; defense-indepth, and the relative risks that have been identified for normal operations and potential 

accident conditions. Consideration should also be given to the design parameters and 

methodology approved in the SAR and their possible use in subsequent 10CFR 72.48(c) 

changes to the design or procedures by the licensee or certificate holder. Any aspect of the 

design or procedures that the NRC determines should not be changed by either the certificate 

holder or general licensee, without prior NRC approval, must be placed in the CoC conditions or 

in the attached technical specifications. 

As described further below, each review procedure is prioritized using a graded approach that 

factored in many of these considerations for a typical review. The prioritization was developed 

with the expertise of NRC reviewers within each discipline, who have several years of regulatory 

experience with the current fleet of certified spent fuel storage cask designs. These priorities 

are intended to serve as a guidepost to the depth and rigor that is expected for a typical review; 

but should not be treated as absolutes for every case. It is the responsibility of the individual 

reviewer to assess the design and determine the ultimate rigor needed to make a safety 

determination, with reasonable assurance, in each review area. In other words, reviewers 

should consistently apply these review procedures for each case, but may need to adjust the 

scope of review in some areas based on safety margins, operational experience, defense-indepth considerations, design novelty, or other issues that are unique to each proposed design. 

Review Process

The purpose of the staff review is to evaluate the proposed cask design, contents and 

operations, and provide regulatory confirmation of reasonable assurance of safe design and 

construction of the cask. 

The reviews are performed by project management and technical review staff with expertise in 

the technical discipline areas described in the review plan. Due to the complexity of the 

technical information in the application, coordination among the different disciplines is important 

to ensure a consistent, uniform, and quality review. As described in the flow charts of each 

chapter, technical issues can overlap between the disciplines and many rely on input from other 

areas. 

This SRP is guidance meant to be used in unison with the current ISGs. ISGs provide guidance 

concerning specific, important issues that either are not currently addressed in the SRP or need 

clarification beyond that in the present SRP text and may delineate specific review procedures. 

For this reason, the staff should be familiar with ISGs that may supersede this guidance and 

these new ISGs should be used together with this SRP in the review of a DSS application. 

ISGs may be discontinued if they are fully incorporated into all applicable regulatory guidance 

documents. Appendix C lists the ISGs from 1 to 22, and identifies which ones have been 

incorporated in this revision of the SRP. 
 3 

The staff may consult the SERs of previous CoC amendments, if reviewing an amendment to a 

currently approved design, as well as the SERs for approved systems of similar design to 

understand past NRC determinations regarding analyses affecting or similar to those in the 

application under review. 

For amendments, the staff should review the entire amendment to ensure that all the licensing 

changes have been identified by the applicant. Amendments may range from minor changes in 

the design, contents, or operations, to adding new major component designs such as storage 

overpacks, transfer casks, and canisters.. Some amendments such as content and design 

changes, are founded upon the design and methodologies previously reviewed by NRC for that 

system. Evaluation of amendment changes to a DSS are often based on the performance of 

the contents, canister, and overpacks as an integrated system. As a result, portions of 

previously approved components, contents, or methodologies in the SAR may be re-examined 

to ensure that the new system under the amendment proposal meets Part 72 requirements. 

During the audit review of an amendment, the staff may occasionally find errors or other safety 

questions that affect part of the previously approved design. The staff may need to review that 

part of the SAR and ask questions to assure the design remains safe and compliant with 

applicable regulations. The questions should be limited to understanding and resolving the 

specific technical issue, and should consider past precedents, regulatory guidance, and risk 

significance, as appropriate. The staff should also consider other processes (e.g. inspections, 

enforcement actions, generic issue program, etc..) to resolve these potential type of safety 

questions with a previously approved design.. 

In case the reviewer finds that the information provided in the SAR is not properly justified, the 

reviewer may develop and then forward to the applicant questions requesting clarification of 

technical issues via a Request for Additional Information (RAI). The applicant’s response to the 

RAI should be reviewed for accuracy as well as the need to update the applicant’s SAR. The 

RAI process is repeated as necessary, consistent with NRC’s in-office instructions, until the 

application is deemed technically acceptable, or until the application review is terminated by the 

NRC or withdrawn by the applicant. 

Once the technical review is complete, a draft SER is written that summarizes the results of the 

review and the cognizant NRC Project Manager approves the SER. If the NRC intends to 

approve the application, the staff prepares Federal Register notices for a direct final rule and a 

companion proposed rule. The rulemaking notices identify the ADAMS numbers for the draft 

CoC, TSs and SER. During the rulemaking process, stakeholders and members of the public 

are allowed to comment on the draft CoC, TSs and SER. After addressing and responding to 

any public comments, the NRC staff modifies the proposed CoC, TS and preliminary SER, if 

necessary, and issues the Final CoC, TS, and SER. The rulemaking adds the CoC, or in the 

case of an amendment to an existing CoC, the CoC amendment, to the list of approved cask 

designs in 10 CFR 72.214. 

Safety Evaluation Report and Content 

The results of a SAR review are documented in an SER. The final determination of the 

organization of an SER is determined by the review project manager, but the SER typically is 

organized in the same manner as this SRP and contains the following information: 

• A general description of the system, operational features, and SNF 

specifications. 
 4 

 • A summary of the approach used by the applicant to demonstrate compliance 

with the regulations, and a description of the reviews that the staff performed to 

confirm compliance. 

 • Comparison of systems, components, analyses, data, or other information 

important in the review analysis to the acceptance criteria, in addition to, 

conclusions regarding the acceptability, suitability, or appropriateness that this 

information provides reasonable assurance the acceptance criteria has been 

met. 

 • Summary of aspects of the review that were selected or emphasized; matters 

that were modified by the applicant: aspects of the cask's design that deviates 

from the criteria stated in the SRP; and the bases for any deviations from the 

SRP. 

 • Summary statements for evaluation findings at the end of each chapter. 

Content of SRP 

Each chapter of the SRP is organized into the following sections: 

 • Review Objective 

 • Areas of Review 

 • Regulatory Requirements 

 • Acceptance Criteria 

 • Review Procedures 

 • Evaluation Findings 

Review Objective. This section provides the purpose and scope of the review and establishes 

the major review objectives for the chapter. The reviewer should obtain reasonable assurance 

during the review that the objectives are met. It also discusses the information needed or 

coordination expected from reviewers of other SAR chapters to complete the subject technical 

review. 

Areas of Review. This section describes the systems, components, analyses, data, or other 

information and their sequence in the discussion of acceptance criteria and review procedures 

sections of each chapter. 

Regulatory Requirements. This section summarizes the regulatory requirements from 

10 CFR Part 72 pertaining to the given SAR section. This list is not all inclusive (e.g., some 

parts of the regulations, such as 10 CFR Part 20, are assumed to apply to all chapters of the 

SAR). 10 CFR Part 72 sections applicable to a DSS are listed in 10 CFR 72.13(d). In addition, 

10 CFR 72.13(c) is important to the applicant to ensure that the general licensee does not 

violate those conditions. The reviewer should read the complete language of the current 

version of 10 CFR Part 72 to determine the proper set of regulations for the section being 

reviewed. 

Acceptance Criteria. This section addresses the design criteria and in some cases specific 

analytical methods that NRC staff reviewers have found to be acceptable for meeting regulatory 

requirements, specified in 10 CFR Part 72, that apply to the given SAR chapter. The 

acceptance criteria are organized in accordance with the review areas established in Section 2 
 5 

of the specific chapter and identify the type and level of information that should be in the 

application. 

These acceptance criteria typically set forth the solutions and approaches that staff reviewers 

have previously determined to be acceptable in addressing a specific safety concern or design 

area that is important to safety. These solutions and approaches are discussed in the SRP so 

that staff reviewers can implement consistent and well-understood positions as similar safety 

issues arise in future cases. These solutions and approaches are acceptable to the staff, but 

they are not the only possible solutions and approaches. 

Substantial staff time and effort has gone into developing these acceptance criteria. 

Consequently, a corresponding amount of time and effort may be required to review and accept 

new or different solutions and approaches. Thus, applicants proposing solutions and 

approaches to new safety issues or analytical techniques other than those described in the SRP 

may experience longer review times and more extensive staff questioning in these areas. An 

alternative for the applicant is to propose new methods on a generic basis, apart from a specific 

license application. Such an alternative proposal could consist of a submittal of a Topical Safety 

Analysis Report (TSAR). This type of application could form the basis for either a change in the 

staff interpretation of the regulatory requirements or support a request for rulemaking to change 

the requirements themselves. 

Review Procedures. 

This section presents a general approach that reviewers typically follow to establish reasonable 

assurance that the applicable acceptance criteria have been met. As an aid to the reviewer, this 

section may also provide information on what has been found acceptable in past reviews. 

Standards that have been found acceptable in specific licensing reviews, or are desirable, but 

not specifically identified in existing regulatory documents, are identified in this section. Since 

many of the reviews are interdisciplinary, the reviewer should coordinate with other reviewers, 

as necessary, for identification of issues in other SAR chapters. 

Each review procedure has been assigned a HIGH, MEDIUM or LOW priority, following 

application of the prioritization process described in Appendix B. These priorities are intended 

to provide guidance to the reviewer regarding the relative level of effort typically applied in 

implementing each procedure. As previously discussed, unique aspects of an application may 

result in an adjustment to the scope of review in a specific technical area. Specifically, the 

following can be used as general guidance on the implications of the priorities for the staff 

review: 

HIGH priority means the NRC staff review should ensure all items in the applicant’s 

submittal are complete and correct as specified in the review procedure. This 

represents the most comprehensive review where many of the analytical methods, 

assumptions, and supporting references are evaluated. The reviewer may need to 

perform independent confirmatory analysis to validate the results of the safety analysis 

calculations. It is expected a reviewer would spend approximately 60 percent of his or 

her review time focused on the high priority review procedures. 

MEDIUM priority means the NRC staff should review the applicant’s submittal for 

completeness and correctness in key areas. This represents a review in which key 

analytical methods, key assumptions, and key supporting references are checked and 
 6 

evaluated. It is expected a reviewer would spend approximately 30 percent of his or her 

review time focused on the medium priority review procedures. 

 LOW priority means the NRC staff review should ensure that the applicant’s submittal 

contains all of the requested information. A limited review of selected portions of the 

application for correctness would be performed. Given its relative significance, the 

reviewer should generally consider the applicant’s analysis to be complete and accurate 

and forego independent confirmation, unless there is a reason to believe otherwise. 

However, if a problem is detected, the reviewer must thoroughly evaluate and resolve 

the issue. It is expected a reviewer would spend approximately 10 percent of his or her 

review time focused on the low priority areas. 

The prioritized review procedures are intended to ensure that staff focus most of their effort on 

the areas considered to have the greatest impact on safety and compliance with regulatory 

limits. While some issues could possibly escape detection and resolution through this audit 

review, they would be of lower regulatory significance. It is important to remember that the 

priority designations were developed on a generic basis and may need to be adjusted 

depending upon the characteristics of specific applications. It is the responsibility of the 

individual reviewer to assess the design and determine the ultimate rigor needed to make a 

safety determination, with reasonable assurance, in each review area. 

Finally it should be noted that a low or medium priority review procedure does not mean an 

application is exempted from any associated regulatory requirement, design requirement, or 

safety analyses that is expected within the review objectives and acceptance criteria in this 

SRP. 

Evaluation Findings. This section provides example summary statements for evaluation 

findings to be incorporated into the SER for each area of review. The evaluation findings are 

prepared by the reviewer based on the satisfaction of the regulatory requirements. The findings 

are published in the SER. 
 1-1 

GENERAL INFORMATION EVALUATION 

1.1 Review Objective 

The purpose of reviewing the general description of the Spent Fuel dry storage system (DSS) is 

to ensure that the applicant has provided a non-proprietary description, or overview, that is 

adequate to familiarize reviewers and other interested parties with the pertinent features of the 

system. 

1.2 Areas of Review 

The general description should be reviewed by all reviewers, regardless of their specific review 

assignments, to obtain a basic understanding of the DSS, its components, and the protections 

afforded for the health and safety of the public. Because much of the information relevant to 

this initial aspect of the DSS review is presented in more detail in other chapters of this SRP, 

this chapter focuses on familiarization with the DSS and consistency of the DSS general 

description with the remaining chapters of the safety analysis report (SAR). The SAR should be 

reviewed for adequacy of the DSS and DSS support system descriptions and drawings. Areas 

of review addressed in this chapter include the following: 

DSS Description and Operational Features

Drawings

DSS Contents

Qualifications of the Applicant

Quality Assurance

Consideration of 10 CFR Part 71 Requirements Regarding Transportation 

1.3 Regulatory Requirements 

This section presents a summary matrix of the portions of U.S. Code of Federal Regulations 

(CFR) Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, 

High-Level Radioactive Waste and Reactor-Related Greater Than Class C Waste,” Title 10, 

“Energy” (10 CFR Part 72) that are relevant to the review areas addressed by this chapter. The 

NRC staff reviewer should read the exact regulatory language. Table 1-1 matches the relevant 

regulatory requirements associated with this chapter to the areas of review. 
 1-2 

Table 1-1 Relationship of Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Regulations

72.2(a)(1), 

(b) 

72.122

(a), 

(h)(1) 

72.140 

(c)(2) 

72.230 

(a) 

72.230

(b) 

72.236(a), 

(c), 

(h),(m) 

DSS Description and 

Operational Features ● ● ! 

Drawings ● ! 

DSS Contents ● ●

Qualifications of the Applicant ● 

Quality Assurance ● ! 

Consideration of 10 CFR 

Part 71 Certified Transportation 

Cask System Requirements 

● ● !

1.4 Acceptance Criteria 

This section identifies the acceptance criteria for the material provided in the introduction. This 

initial aspect of the DSS review should contain sufficient information to allow all reviewers, 

regardless of their specific review assignments, to understand the principal functions and design 

features of the DSS. 

1.4.1 DSS Description and Operational Features 

The application should contain a broad overview and a general, non-proprietary description 

(including engineering drawings, sketches, and illustrations) of the DSS. This information 

should clearly identify the functions of all principal components and principal auxiliary 

equipment, and provide a list of those components classified as being “important to safety.” 

Important aspects from all of the disciplinary areas should be summarized. If there are several 

versions of the cask because of design limitations of nuclear power plants and ISFSIs, the 

differences between the versions should be delineated. Typical operational sequences for 

loading and unloading procedures should be described. 

If the potential exists that the DSS will be used to store damaged fuel, the SAR should include a 

discussion of how the sub-criticality requirement of 10 CFR 72.236(c) and the wet or dry loading 

and unloading requirements of 10 CFR 72.236(h) will be maintained. 

The reviewer should verify that any documents submitted to the NRC in other applications and 

incorporated in whole or in part have been indexed, and a summary has been included in the 

appropriate section of the SAR. 
 1-3 

1.4.2 Drawings 

Drawings should be included in the first chapter of the SAR. The drawings should contain 

sufficient detail to allow the reviewer to understand the operation of the DSS and any special 

equipment used for loading, unloading, transportation, or long-term storage of the DSS. Also, 

the drawings should provide enough detail to allow the reviewer the option of developing an 

analysis model for confirmatory calculations. 

Ideally, the drawings should be non-proprietary. However, in some cases, the applicant may 

request to have certain specific portions of the drawings classified as proprietary. Reviewers 

should note that any drawings relied on as the technical basis for adding the DSS design to the 

list of approved DSSs contained in Subpart K of 10 CFR 72 become part of the public record. 

Such drawings will not be treated as proprietary and will be made available to the public 

[10 CFR 2.390]. 

Any request for withholding from public disclosure subject to the provisions of 10 CFR 2.390 

should be accompanied by an affidavit and must include information to support the claim that 

the material is proprietary. The NRC Project Manager will develop and administer public 

disclosure determinations, and the Office of the General Counsel will review them for 

compliance with the requirements of 10 CFR 2.390. 

1.4.3 DSS Contents 

The reviewer should ensure specifications are provided for the contents expected to be stored 

in the DSS (normally spent nuclear fuel [SNF]). These specifications may include, but not be 

limited to, type of SNF (i.e., boiling-water reactor [BWR], pressurized-water reactor [PWR], or 

both); number of SNF assemblies the cask can accommodate; maximum allowable enrichment 

of the fuel before any irradiation; burnup (i.e., MWd/MTU); minimum acceptable cooling time of 

the SNF before storage in the DSS (e.g., aged at least 1 year); maximum heat designed to be 

dissipated; maximum SNF loading limit; condition of the SNF (i.e., intact, undamaged, 

damaged, etc.); weight and nature of non-SNF contents; and inert atmosphere requirements. 

1.4.4 Quality Assurance 

Reviewers should verify that the application describes the proposed quality assurance (QA) 

program and cites the applicable implementing procedures. This description should satisfy all 

requirements of 10 CFR Part 72, Subpart G. A detailed review of the QA program to be 

described in the SAR is presented in Chapter 14, “Quality Assurance Evaluation,” of this SRP. 

1.4.5 Consideration of 10 CFR Part 71 Requirements Regarding Transportation 

If the DSS has previously been evaluated for use as a transportation cask, the submittal should 

include the Part 71 Certificate of Compliance (CoC) and associated documents in accordance 

with 10 CFR 72.230(b). If application for storage is submitted, the transportability, per 10 CFR 

72.236(m) should be addressed. (See Section 1.5.5). 

1.5 Review Procedures 

Figure 1-1 presents an overview of the evaluation process and a complete bulleted listing of 

pertinent information for each chapter. Figure 1-1 and the corresponding figures in each 
 1-4 

chapter of this Standard Review Plan (SRP) provide a means to coordinate the review among 

the NRC staff disciplines. 

Regulatory requirements of 10 CFR Part 72 applicable to the general description review are 

delineated in the following subsections. Since the review of the General Description of the SAR 

is interdisciplinary, the reviewer should coordinate with other reviewers (e.g., structural, thermal, 

shielding, criticality, materials), as necessary, for identification of related issues. 

1.5.1 DSS Description and Operational Features (MEDIUM Priority)

Reviewers should verify that the application provides a broad overview of the DSS design that is 

non-proprietary and may be used as a tool to familiarize interested parties with the features of 

the proposed DSS. This description should present the principal characteristics of the DSS 

including its dimensions, weight, and construction materials. In addition, the description should 

clearly identify all components considered important to safety. Features such as the 

confinement vessel, fuel basket, valves, lids, seals, penetrations, trunnions, closure 

mechanisms, shielding safety features, criticality control features, impact limiters, and cask 

identification should be identified and described. A clear definition of the primary confinement 

system is particularly important. Special design features of the DSS such as a non-passive 

heat-removal system, neutron poisons or monitoring instrumentation should be discussed. 

Sketches and diagrams found throughout the SAR should be compared with the detailed 

drawings presented in SAR Chapter 1, “General Information”. If the application includes 

proprietary drawings and descriptions that will remain proprietary upon approval of the license 

or certificate, the sketches, drawings, and diagrams that provide the general description and 

operational features need not show the proprietary features. This may be achieved by depicting 

less detail or by illustrating generic components that fulfill the design function. However, these 

representations should show the operational concept and features important to safety in 

sufficient detail to form an acceptable basis for public review and comment. 

In addition to information on a single DSS, the application should describe any limitations on the 

arrangement of DSS arrays. For a particular DSS, these limitations may include the minimum 

spacing between the casks, maximum density of casks in an array, and/or total number of casks 

or amount of SNF that may be stored at a single ISFSI. The acceptable limitations should be 

included among the technical specifications in the Safety Evaluation Report (SER) (see Chapter 

13, “Technical Specifications and Operating Controls and Limits Evaluation,” of this SRP). For a 

DSS such as those with metal confinement vessels stored in a concrete vault, information on 

the configuration of vault compartments and horizontal/vertical arrangement is necessary. The 

operational sequences for loading and unloading the cask should be described. 

Damaged fuel may require canning for storage and transportation. The purpose of canning is to 

confine gross fuel particles to a known, subcritical volume during off-normal and accident 

conditions, and to facilitate handling and ready retrieval of contents. Therefore, the reviewer 

should verify that a description of how damaged fuel would be canned, the characteristics of the 

can, and the means in which the can would be placed in the cask and either readily retrieved 

(recovered) or retrieved is in the application. 
 1-5 

Figure 1-1 Overview of Safety Evaluation 

Compliance with 10 CFR Part 72

• Condition of DSS after Testing per Applicable 

Portions of §72.122

• Structural, Thermal, Confinement, Shielding, 

Criticality Requirements, and Materials

• Operating Procedures, Acceptance Tests, 

and Maintenance

DSS Description

• Cask and Overpack

• Operating Features

• Contents of DSS

DSS Design Information

• Purpose of Application

• Quality Assurance Program

• Proposed Use and Contents of DSS

• DSS Category, Type, and Model Number

• Thermal Loading

• Fabrication and Welding Criteria

Chapter 1 – General Information Evaluation

• Spent Fuel Design Basis

• External Conditions

• Items Important to Safety

• DSS Design Basis

Chapter 2 – Principal Design Criteria Evaluation

• Component Materials

• Dimensions and Weights

• Internal and External Structure

• Codes and Standards

Chapter 3 – Structural Evaluation

• Decay Heat

• Heat Dissipation

• Component Materials

• Dimensions

• Spent Fuel Cladding

• Configuration

Chapter 4 – Thermal Evaluation

• Allowable Leak Rate • Penetrations

• Accident Conditions

• Containment Boundary

• DSS Contents

• Dimensions

• Component Materials

Chapter 5 – Containment Evaluation

• Component Materials

• Content Limits

• Dimensions

• Configuration

Chapter 6 – Shielding Evaluation

• Component Materials

• Neutron Poison Contents

• Fissile Content Materials

• Dimensions and Tolerance

Chapter 7 – Criticality Evaluation

• Material Selection • Corrosion • Cladding Integrity

Chapter 8 – Materials Evaluation

• General Restrictions

• Operational Sequences

Chapter 9 – Operating Procedures Evaluation

• Maintenance Tasks

• Instrumentation

• Fabrication Materials

• Contents

• Codes and Standards

• Dimensions and Tolerances

Chapter 10 – Acceptance Tests and Maintenance Program Evaluation

• Dose Assessment

• Health Physics Program

• ALARA

• Radiation Protection Features

Chapter 11– Radiation Protection Evaluation

• Accident Identification • Corrective Action Program

• DSS Performance Analysis

Chapter 12 – Accident Analysis Evaluation

• Transport

• Surveillance

• Operating, Monitoring, and Safety Limits

• Loading and Unloading

Chapter 13 – Technical Specifications and Operating Controls Evaluation

• Items Important to Safety

• Document Control

• Program Description

• National Standards

Chapter 14 – Quality Assurance Evaluation

Figure 1-1 Overview of Safety Evaluation
 1-6 

1.5.2 Drawings (MEDIUM Priority) 

Drawings are usually presented in Chapter 1, “General Information” of the SAR. Reviewers 

should be familiar with NUREG/CR-5502, “Engineering Drawings for 10 CFR Part 71 Package 

Approval.” While NUREG/CR-5502 was written for transportation packages, the criteria in 

NUREG/CR-5502 for drawings can be applied to applications for storage casks. 

Although some applications may contain drawings designated as “proprietary,” reviewers should 

note that any drawings relied on as the technical basis for adding the DSS design to the “list of 

approved spent-fuel storage DSS” contained in Subpart K of 10 CFR 72 become part of the 

public record. Such drawings will not be treated as proprietary and will be made available to the 

public [10 CFR 2.390(a)]. Applicants may submit additional drawings showing greater detail to 

support their evaluations, and these may be exempted from the public record if they are not 

relied on by the staff as part of the technical basis for DSS design approval. The reviewer 

should verify that all structures, systems, and components (SSC) important to safety are 

sufficiently detailed to enable reviewers to evaluate their effectiveness. In addition, information 

on non-safety items may also be necessary to ensure they do not impede the safety systems. 

Each reviewer should evaluate the level of detail furnished with the application. The drawings 

should specify those details of the cask design that affect its evaluation. Those design features 

that have a significant effect on safety if altered or modified, should be considered for inclusion 

into the technical specifications directly or by reference. If size reduction has rendered any 

information unclear or illegible, the Project Manager in the Division of Spent Fuel Storage and 

Transportation (SFST) should request that the applicant provide larger or full-size drawings. 

Particular attention should be devoted to ensuring that dimensions, materials, and other details 

on the drawings are consistent with those described in both the text of the SAR and those used 

in supplementary analysis. The dimensions shown on the general arrangement drawing should 

specify the overall size of the cask and the location and configuration of the contents. All 

dimensions indicated on drawings should include tolerances that are consistent with the cask 

evaluation. 

1.5.3 DSS Contents (MEDIUM Priority)

The application should present a general description of the contents proposed for storage in the 

DSS. Because a very detailed description of the proposed DSS contents or SNF is typically 

provided in Chapter 2, “Principal Design Criteria,” of the SAR, the information presented in 

Chapter 1, “General Information” of the SAR is important only to the extent that it permits overall 

familiarization with the DSS. Key parameters for SNF include the type of fuel (i.e., PWR, BWR, 

or both), number of fuel assemblies, the radiation source terms associated with these fuel 

assemblies, preferential loading, and condition of the fuel assemblies (i.e., intact or 

consolidated). Chapter 1 may also include additional characteristics such as maximum burnup, 

initial enrichment, heat load, and cooling time as well as the assembly vendor and configuration 

(e.g., Westinghouse 17x17). These characteristics may also be repeated in Chapter 2. In 

addition, the cover gas, if any, should be identified. 

If the applicant proposes the storage of damaged fuel or components that are associated with or 

integral to the fuel assembly that do not have an integral confinement boundary, the range of 

permissible conditions for the stored material should be defined. If the DSS system is intended 

to be used to store damaged fuel or components that are associated with or integral to the fuel 

assembly with an integral confinement boundary when placed in the confinement DSS, the 
 1-7 

possible range of conditions of the fuel or components should be stated. 10 CFR 72.122(h)(1) 

requires “canning” or use of other acceptable means for storing fuel with cladding that is not or 

may not remain intact and for unconsolidated assemblies (without intact cladding). 

10 CFR 72.236(c) requires the damaged fuel be maintained in a subcritical condition, while 

10 CFR 72.236(h) requires the damaged fuel to be compatible with wet or dry loading and 

unloading facilities. If damaged fuel is to be stored, the application should address how the 

following basic requirements will be met: 

 • Maintain subcriticality; 

 • Prevent unacceptable release of contained radioactive material; 

 • Avoid excessive radiation dose rates and doses; 

 • Maintain ready retrieval of the contents. 

If the application requests approval to use the DSS system to store components that are 

associated with or integral to the fuel assembly (i.e., control spiders, burnable poison rod 

assemblies, control rod elements, thimble plugs, fission chambers, and primary and secondary 

neutron sources, or BWR channels that are an integral part of the fuel assembly that do not 

require special handling), the application should present summary descriptions of those 

components in Chapter 1, “General Information” of the SAR. The SFST staff has made a 

practice of carefully characterizing components as being “associated with or integral to” the fuel 

assembly because only those components listed above are acceptable at a geologic repository 

per 10 CFR 961.11, Appendix E, Section B.2. Components that are associated with or integral 

to the fuel assembly are reviewed in more detail as part of Chapter 2, “Principal Design Criteria 

Evaluation,” of this SRP. Also, if the components are degraded (e.g., the component does not 

provide adequate confinement under design basis conditions to contain radioactive gas or other 

dispersible radioactive materials), the application should describe the possible conditions and 

alternative confinement methods, if any. 

1.5.4 Quality Assurance Program (See Chapter 14 for Priority)

The application should describe the proposed QA program, citing all implementing procedures 

in a manner that satisfies the 18 criteria defined in 10 CFR Part 72, Subpart G, “Quality 

Assurance” (10 CFR §§ 72.142-72.176). The description need only refer to procedures that 

implement the QA program, and these procedures need not be explicitly included in the 

application. The QA program should address design, fabrication, construction, testing, 

operation, and modification activities regarding the SSCs that are important to safety. The 

application should also discuss the activities to be performed under the QA program and how 

these activities will be controlled to ensure compliance with all of the requirements of Subpart G. 

These controls may be applied to the various activities using a graded approach as presented in 

NUREG/CR-6407, “Classification of Transportation Packaging and Dry Spent Fuel Storage 

System Components According to Importance to Safety” (i.e., QA efforts expended for a given 

activity should be consistent with that activity's system classification and function). 

Per 10 CFR 72.140(d), a QA program previously approved by the NRC and established, 

maintained, and executed for another DSS will be accepted as satisfying the requirements for a 

QA program for the purpose of this application. Additionally, previously approved QA programs 

that meet the requirements of Appendix B to 10CFR 50 or Subpart H to 10 CFR 71, will be 

acceptable provided they also meet the recordkeeping requirements of §72.174. Any reference 

to a previously approved QA program should identify the program by date of submittal to the 

NRC, docket number, and date of NRC approval. The reviewer should coordinate with the 

Chapter 14, “Quality Assurance Evaluation,” review of this SRP. 
 1-8 

1.5.5 Consideration of 10 CFR Part 71 Requirements (MEDIUM Priority)

Casks that have been certified for transportation of SNF under 10 CFR Part 71 may be 

approved for the storage of SNF under 10 CFR Part 72 provided the application contains: 

 • A copy of the CoC issued under 10 CFR Part 71, 

 • Copies of all drawings and other documents referenced in the 10 CFR Part 71 

CoC, and 

 • Sufficient information in the SAR to demonstrate that the cask is suitable for the 

storage period of SNF as defined by 10 CFR 72.230(b). 

Because applications for dual-purpose certification under 10 CFR Parts 71 and 72 are 

sometimes submitted jointly, the final (approved) version of such documents may not be 

available at the time of initial DSS SAR submission. Nonetheless, applicable documentation of 

the Part 71 certification (or application), including questions and responses from the related 

review, should be provided to the Part 72 review team, as appropriate. 

Substantial coordination of the Part 71 and Part 72 reviews is necessary to ensure consistency 

and avoid duplication of effort. The reviewer should verify that a process for promptly informing 

each of the review teams about DSS system design changes precipitated by any concurrent 

safety reviews has been identified by the applicant. Provisions for communicating these 

changes should be addressed by, and discussed with, the applicant. In addition, transportability 

of storage-only or dual purpose casks, per 10 CFR 72.236(m) should be addressed. The 

applicant should address how it is planning to address the transportation requirements. The 

reviewer should verify that such considerations have been made and described in the SAR, 

when the SAR and/or accompanying documentation indicate plans to use the cask system for 

transportation purposes. 

1.6 Evaluation Findings 

The evaluation findings are prepared by the reviewer on satisfaction of the regulatory 

requirements in Section 1.3. These statements should be similar to the following examples, if 

the documentation submitted with the application supports positive findings for each of the 

regulatory requirements (the finding number is for convenience in reference within the SRP and 

SER): 

 F1.1 A general description and discussion of the DSS is presented in Section(s) 

of the SAR, with special attention to design and operating characteristics, 

unusual or novel design features, and principal considerations important to 

safety. 

 F1.2 Drawings for SSCs important to safety are presented in Section of the 

SAR. A listing of those drawings (including dates and revision numbers) that 

were relied upon as a basis for approval appears in Section of the SER. 

 F1.3 Specifications for the SNF to be stored in the DSS are provided in SAR 

Section ___ . Additional details concerning these specifications are presented 

in Chapter of both the SAR and SER. 
 1-9 

 

 F1.4 The quality assurance program and implementing procedures are described in 

Section of the SAR. 

 F1.5 The [DSS system designation] [has been/is/is not being] certified under 10 CFR 

Part 71 for use in transportation. A copy of the SAR and CoC issued under 

10 CFR Part 71 is on file with the NRC under Docket No. [if applicable]. 

A summary statement similar to the following should be made: 

“The staff concludes that the information presented in Chapter 1, “General Information” 

of the SAR satisfies the requirements for the general description under 10 CFR Part 72. 

This finding is reached on the basis of a review that considered the regulation itself, 

Regulatory Guide 3.61, and accepted practices.” 
 2-1 

2 PRINCIPAL DESIGN CRITERIA EVALUATION 

2.1 Review Objective 

The objective of evaluating the principal design criteria related to structures, systems, and 

components (SSCs) important to safety is to ensure that, in the view of the U.S. Nuclear 

Regulatory Commission (NRC) staff, the principal design criteria comply with the relevant 

general criteria established in U.S. Code of Federal Regulations (CFR) Part 72, “Licensing 

Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive 

Waste and Reactor-Related Greater Than Class C Waste,” Title 10, “Energy” (10 CFR Part 72). 

Further guidance can be found in NUREG/CR-6407, “Classification of Transportation Packaging 

and Dry Spent Fuel Storage System Components According to Importance to Safety.” Material 

provided in this chapter will form the basis for accepting the safety analysis report (SAR) for 

NRC staff review. 

With regard to reviewing the principal design criteria, the applicant may take one of two 

approaches: (1) SAR Chapter 2, “Principal Design Criteria” may discuss these criteria in general 

terms with details provided in later sections or (2) SAR Chapter 2 may present detailed 

discussions of selected (or all) criteria. Past applicants have generally selected the latter 

approach. Subsequent chapters of this Standard Review Plan (SRP) provide detailed 

discussions of the design criteria applicable to each functional area (e.g., structural, thermal) 

without regard to those that may have been presented in SAR Chapter 2. 

2.2 Areas of Review 

The review of the principal design criteria should provide reasonable assurance that all design 

criteria are addressed in the SAR. The following areas of review have been adopted by the 

NRC staff: 

Structures, Systems, and Components Important to Safety

Design Basis for Structures, Systems, and Components Important to Safety

Spent Nuclear Fuel (SNF) Specifications 

 External Conditions 

Design Criteria for Safety Protection Systems

 General 

 Structural 

 Thermal 

 Shielding/Confinement/Radiation Protection 

 Criticality 

 Material Selection 

 Operating Procedures 

 Acceptance Tests and Maintenance 

 Decommissioning 

2.3 Regulatory Requirements 

This section presents a summary matrix of the portions of U.S. Code of Federal Regulations 

(CFR) Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, 

High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste” Title 10, 
 2-2 

“Energy” (10 CFR Part 72) that are relevant to the review areas addressed by this chapter. The 

NRC staff reviewer should read the exact regulatory language. Table 2-1 matches the relevant 

regulatory requirements associated with this chapter to the areas of review. 

Table 2-1 Relationship of 10 CFR Part 72 Regulations and Areas of Review 

Areas of 

Review

10 CFR Part 72 Regulations

72.2 

(a)(1) 

72.104 

(a), 

(b), 

 (c) 

72.106 

(a), 

(b), 

 (c) 

72.122 

(a), (b) 

(1)(2) 

(3), 

(c), (f) 

72.122 

(h)(1) 

(4) 

72.122 

(i), (l) 

72.124 

(a), (b) 

72.126 

(a)(1) 

(2)(3) 

(4)(5) 

(6) 

72.236 

(a), 

(b), 

(c), (d) 

72.236 

(e), (f), 

(g), 

(h), (i), 

(l), (m) 

SSCs 

Important to 

Safety 

 ● 

Design 

Bases for 

SSCs 

Important to 

Safety 

● ● ● 

Design 

Criteria for 

Safety 

Protection 

Systems 

● ● ● ● ● ● ● ● ●

2.4 Acceptance Criteria 

The reviewer should verify that the applicant has provided either sufficient general or summary 

discussions of the SSC design features and of both operational and accident conditions. This 

demonstrates a clear and defensible case that they have met the design criteria. In evaluating 

the principal design criteria related to DSS SSCs that are important to safety, reviewers should 

seek to ensure that the given design fulfills the following acceptance criteria. 

2.4.1 SSCs Important to Safety 

The reviewer should verify that the applicant presents the general configuration of the DSS and 

provides an overview of specific components and their intended functions. In addition, the 

reviewer should ensure the applicant identifies those components deemed to be important to 

safety and addresses the safety functions of these components in terms of how they meet the 

general design criteria and regulatory requirements discussed above. Additional information 

concerning specific functional requirements for individual DSS components is addressed in 

subsequent chapters of this SRP. 
 2-3 

2.4.2 Design Bases for SSCs Important to Safety 

Detailed descriptions of each of the items listed below are generally found in specific sections of 

the SAR. However, a brief description of these areas, including a summary of the analytical 

techniques used in the design process, should also be captured in Chapter 2, “Principal Design 

Criteria” of the SAR. This description gives reviewers a perspective of how specific DSS 

components interact to meet the regulatory requirements of 10 CFR Part 72. This discussion 

should be non-proprietary since it may be used to familiarize interested persons with the design 

features and bounding conditions of operation of a given DSS. 

2.4.2.1 SNF Specifications 

The range and types of SNF or other radioactive materials that the DSS is designed to store 

should be specified. In addition, these specifications should include, but are not limited to: 

 • The type of SNF (i.e., boiling-water reactor (BWR), pressurized-water reactor 

(PWR), or both), 

 

 • Cladding material, 

 • Maximum assembly uranium mass loading, 

 • Weights of the stored materials, 

 • Dimensions and configurations of the fuel, 

 • The identification and limits on amount and position of damaged fuel, if damaged 

fuel is to be stored, and the dimensions of the “damaged-fuel can,” 

 • Maximum allowable enrichment of the fuel before any irradiation for criticality 

safety and minimum enrichment for the shielding evaluation, 

 • Assigned Burnup Loading Value (i.e., MWd/MTU), 

 • Loading Curves for each set of licensing conditions if Burnup Credit is used 

(required minimum burnup versus enrichment curve), 

 • Operational history parameters (e.g., average in-core soluble boron 

concentration, average moderator temperature, etc.) if burnup credit is used, 

 • Minimum acceptable cooling time of the SNF before storage in the DSS, 

 • Maximum heat to be dissipated, 

 • Maximum number of SNF elements, 

 • Condition of the SNF (i.e., intact assembly, damaged fuel or consolidated fuel 

rods), 

 Inerting atmosphere requirements and the maximum amount of fuel permitted for 

storage in the DSS. 
 2-4 

For DSSs that will be used to store components that are associated with or integral to fuel 

assemblies (e.g., control rods and BWR fuel channels), the reviewer should ensure the 

applicant specifies the types and amounts of radionuclides, heat generation, and the relevant 

source strengths and radiation energy spectra permitted for storage in the DSS. For other 

radioactive materials to be stored with the SNF assemblies, the SAR should specify the 

following: 

 • The design basis source term; 

 • The effects of gas generation on the cask internal pressure; 

 • The effects of the additional weight and length of the proposed material on 

structural and stability analyses; 

 • The impact of the added heat from these components, including the impact on 

heat transfer characteristics; and 

 • Credit for any negative reactivity from residual neutron absorbing material 

remaining in the control components. 

2.4.2.2 External Conditions 

The SAR should define the bounding conditions under which the DSS is expected to operate. 

Such conditions include both normal and off-normal environmental conditions as well as 

accident conditions. In addition, the reviewer should verify that the applicant has considered the 

effects of natural events such as tornadoes, earthquakes, floods, and lightning strikes. 

2.4.3 Design Criteria for Safety Protection Systems 

2.4.3.1 General 

The SAR should define an expected lifetime for the cask design. The minimum licensing period 

is defined in 10 CFR 72.230(b). The reviewer should verify that the applicant has provided a 

brief description of the proposed quality assurance (QA) program, and applicable industry codes 

and standards, which will be applied to the design, fabrication, construction, and operation of 

the DSS. The applicant should also describe how the cask design reflects consideration of 

compatibility with removal from a reactor site, transportation, and ultimate disposition of the 

stored spent fuel. 

In establishing normal and off-normal conditions applicable to the design criteria for DSS 

designs, applicants should account for actual facility operating conditions. Therefore, design 

considerations should reflect normal operational ranges, including any seasonal variations or 

effects. 

2.4.3.2 Structural 

The SAR should define how the DSS structural components are designed to accommodate 

combined normal, off-normal, and accident loads while preserving recover and protecting the 

DSS contents from significant structural degradation, criticality, and loss of confinement. This 

discussion is generally a summary of the analytical techniques and calculation results from the 
 2-5 

detailed analysis discussed in SAR Chapter 3, “Structural Evaluation,” and should be presented 

in a non-proprietary form. 

2.4.3.3 Thermal 

The SAR should contain a general discussion of the proposed heat-removal systems, including 

the reliability and verifiability of such systems, and any associated limitations. All heat-removal 

systems should be passive and independent of intervening actions under normal and off-normal 

conditions. 

2.4.3.4 Shielding/Confinement/Radiation Protection 

The reviewer should ensure that the applicant describes those features of the cask that protect 

occupational workers and members of the public against direct radiation dosages and releases 

of radioactive material, and minimize the dose after any off-normal or accident-level conditions. 

2.4.3.5 Criticality 

The SAR should address the mechanisms and design features that enable the DSS to maintain 

SNF in a subcritical condition under normal, off-normal, and accident-level conditions. 

2.4.3.6 Material Selection 

The materials selected for the DSS must demonstrate adequate corrosion performance during 

normal operation, off-normal, and accident-level conditions in the environmental conditions of 

the ISFSI for the duration of the license. 

The spent fuel cladding must be protected during storage against degradation that leads to 

gross ruptures, or the fuel must be otherwise confined such that degradation of the fuel during 

storage will not pose operational problems with respect to its removal from storage. 

2.4.3.7 Operating Procedures 

The reviewer should ensure that the applicant provides potential licensees with guidance 

regarding the content of normal, off-normal, and accident response procedures. Cautions 

regarding both loading, unloading, and other important procedures should be mentioned here. 

Retrievability should be provided for normal and off-normal conditions. Applicants may choose 

to provide model procedures to be used as aids in preparing detailed site-specific procedures. 

2.4.3.8 Acceptance Tests and Maintenance 

The reviewer should verify that the applicant identifies the general commitments and industry 

codes and standards used to derive acceptance, maintenance, and periodic surveillance tests 

used to verify the capability of DSS components to perform their designated functions. In 

addition, the reviewer should ensure the applicant discusses the methods used to assess the 

need for such tests with regard to specific components. 
 2-6 

2.4.3.9 Decommissioning 

Casks should be designed for ease of decontamination and eventual decommissioning. The 

reviewer should examine the SAR to ensure the applicant describes the features of the design 

that support these two activities. 

2.5 Review Procedures 

Chapter 2, “Principal Design Criteria” applies to all review disciplines. Figure 2-1 presents an 

overview of the evaluation process and may be used as a guide for the coordination of the 

review among review disciplines. 

Reviewers for each section of the SAR should consider SAR Chapter 2 in combination with 

additional details presented later in the SAR. In this SRP, evaluations of design criteria 

applicable to each of the relevant chapters of the SAR are discussed in detail. Reviewers 

should coordinate the review of each chapter with the applicable disciplines to ensure that multidisciplinary issues, which impact more than one chapter, have been addressed. 

2.5.1 SSCs Important to Safety (MEDIUM Priority) 

Reviewers should verify that the applicant has clearly identified all SSCs important to safety 

(see Glossary for the definition of “important to safety”) and documented the rationale for this 

designation. Such information may be provided in tabular form. Reviewers should review the 

general DSS description presented in SAR Chapter 1, “General Description.” Reviewers should 

ensure that the applicant has provided adequate justification for excluded SSCs. 

Reviewers should pay particular attention to instrumentation and other equipment (e.g., lifting 

devices and transport vehicles). In general, the NRC staff accepts that monitoring systems 

need not be classified as being important to safety. For example, a failure in the functioning of 

the pressure monitoring system does not directly result in a release of radionuclides. Additional 

justification for not considering such systems as being important to safety may be presented in 

later sections of the SAR and summarized in SAR Chapter 2, “Principal Design Criteria”. 

Reviewers should consider adding to SAR Chapter 13 “Technical Specifications and Operating 

Controls and Limits” any design features that would have a significant effect on safety if altered 

or modified. Any such additions to Chapter 13 should be thoroughly discussed in their 

respective sections of the SER. 

2.5.2 Design Bases for SSCs Important to Safety 

The reviewer should verify that the applicant’s design basis for DSS approval accurately 

identifies the range of SNF configurations and characteristics, the enveloping conditions of use, 

the bounding site characteristics, and is consistent with or bounds the DSS’s Technical 

Specifications. These factors determine the bounds within which an ISFSI owner may use the 

SAR rather than provide additional proof regarding suitability of the covered topics. 
 2-7 

Figure 2-1 Overview of Principal Design Criteria Evaluation 

Chapter 1 – General Information 

Evaluation

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation

• DSS Description

• Spent Fuel Storage Specifications

• Items Important to Safety

• External Conditions

• Inspections

• Documentation

• Surveillance

• Test Procedures

• Codes and Standards

• ALARA Requirements

Chapter 11 – Radiation 

Protection Evaluation

Chapter 12 – Accident 

Analysis Evaluation

• Accident Conditions

• Natural Phenomena 

Events

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation

• Components/Features

• Design Basis Source Term

• Monitoring Instrumentation

• Codes and Standards

Chapter 6 – Shielding 

Evaluation

• Sources

• Dose vs. Distance

• Personnel Exposures

• Site Boundary Doses

Chapter 7 – Criticality 

Evaluation

• Double Contingency

• Benchmarking

• Burnup Credit

Chapter 8 – Materials 

Evaluation

• Material Selection

• Corrosion

• Cladding Integrity

Chapter 9 – Operating 

Procedures Evaluation

• Loading/Unloading

• Misloading

• Monitoring

• Retrievability

Chapter 3 – Structural Evaluation

• Loading

• Analysis Methods

• Spent Fuel Specifications

Chapter 4 – Thermal Evaluation

• Heat Load

• Fuel Cladding Limits

• Pressures

• Fire Accident

Chapter 5 – Confinement 

Evaluation

• General Design Criteria

• Spent Fuel Specifications

• Structural Design Criteria

• Thermal Design Criteria

• Shielding/Radiation Protection 

Design Criteria

• Operating Procedures

Design Criteria for Safety Protection Systems

• General

• Structural

• Thermal

• Shielding/Confinement/Radiation Protection

• Criticality

• Materials

• Operating Procedures

• Acceptance Tests and Maintenance

• Decommissioning

Structures, Systems, and Components Important to Safety

Design Basis for Structures, Systems, and Components 

Important to Safety

• Spent Fuel Specification

• External Conditions

Chapter 2 – Principal Design Criteria Evaluation

Design Criteria for Safety Protection Systems

• General

• Structural

• Thermal

• Shielding/Confinement/Radiation Protection

• Criticality

• Materials

• Operating Procedures

• Acceptance Tests and Maintenance

• Decommissioning

Structures, Systems, and Components Important to Safety

Design Basis for Structures, Systems, and Components 

Important to Safety

• Spent Fuel Specification

• External Conditions

Chapter 2 – Principal Design Criteria Evaluation
 2-8 

2.5.2.1 SNF Specifications (MEDIUM Priority) 

The reviewer should review the detailed specifications for the SNF to be stored in the DSS as 

presented in SAR Chapter 2, “Principal Design Criteria” and ensure that they are consistent with 

those specifications discussed in SAR Chapter 1, “General Information” and later in the SAR. 

The description of the range of SNF to be stored should include the type (PWR, BWR, or both); 

configuration (e.g., 17x17, 15x15, or 8x8); fuel vendor; number of assemblies per cask; 

enrichment; burnup and burnup profiles; minimum cooling time; decay heat generation rate; 

type of cladding; physical dimensions; total weight per assembly; and uranium weight per 

assembly. In addition, if components associated with fuel assemblies (e.g., control assemblies) 

will be stored with the fuel, the reviewer should ensure that combined weight, dimensions, heat 

load, and other appropriate information (e.g., number per cask) are specified. 

The reviewer should examine any limitations regarding the condition of the SNF. If damaged 

fuel is allowed, the effects of such damage should be assessed in later sections of the SAR. 

Specific conditions that define damaged fuel are provided in Section 8.6, “Supplemental 

Information for Methods for Classifying Fuel,” of this SRP with methods for classifying fuel 

identified in Section 8.4.17.2 of this SRP. If damaged rods have been removed from a fuel 

assembly, and they have/have not been replaced with solid dummy rods, the criticality reviewer 

should determine whether the intended loading configuration has been adequately analyzed to 

show sub-criticality. Note, the presence of additional moderating material will need to be 

addressed in the criticality analysis in SAR Chapter 7, “Criticality”. Coordination with the 

structural reviewer is necessary if there are structural defects in the assembly hardware. 

The release of fill and fission product gases from failed fuel rods increases the pressure in the 

cask cavity and the potential source term in the event of confinement failure. Consequently, the 

reviewer should verify that the applicant provides information regarding the fill/fission product 

gas present in the fuel as well as the free volume in the cask cavity to enable reviewers to 

evaluate the pressure in the cask cavity resulting from cladding failure during storage. For the 

purpose of calculating internal cask pressures, the NRC staff has accepted the bounding 

assumptions given in SRP Section 4.5.4.6, “Pressure Analysis” regarding the minimum 

percentages of fuel rods have failed (and released their gases). 

The reviewer should pay particular attention to the specification of burnup, cooling time, and 

decay heat generation rate. These parameters are generally not independent, and the manner 

in which they are specified and combined can significantly affect the maximum allowed cladding 

temperature as discussed in SRP Chapter 4, “Thermal Evaluation.” 

The SAR will typically list various fuel assemblies that can be stored in the DSS. It is not 

expected that one type of fuel assembly will be bounding for all analyses. The reviewer should 

ensure that the applicant has justified which specifications are bounding for each of the 

evaluations presented in subsequent sections of the SAR. Specifications used in these 

analyses should also be clearly identified or referenced in SAR 13, “Technical Specifications 

and Operational Controls and Limits”. 

If the applicant requests permission for the storage of components that are associated with or 

integral to the fuel assembly in the cask, the reviewer should examine the relevant detailed 

specifications, conditions, and constraints presented in the SAR. These specifications should 

be as detailed as the applicable information presented for the fuel designs to provide the 

reviewer with a basis for determining that the relevant safety functions of the DSS will be 
 2-9 

maintained. The reviewer should ensure that the applicant also considers the storage of these 

components in the analyses. 

If the applicant requests burnup credit, the reviewer should examine the relevant detailed 

specifications of the contents to which burnup credit is being applied. These specifications 

include those that are already considered in criticality analyses for fresh fuel (e.g., maximum 

initial enrichment). Additional specifications that must be reviewed include the cooling time, the 

burnup, the requested amount of credit (i.e., the specific actinides), and operational history 

parameters (e.g., core average boron concentration and assembly average moderator 

temperature). 

2.5.2.2 External Conditions (MEDIUM Priority) 

The SAR should identify those external conditions that significantly affect, or could potentially 

affect, the performance of the DSS. These design-basis conditions will generally restrict either 

the sites at which the DSS can be used for SNF storage or the manner in which the DSS can be 

handled. For example, by selecting the design earthquake, the SAR limits the use of the cask 

being reviewed to sites that are bounded by this seismic limit. By establishing a design-basis 

drop, the SAR defines the maximum height to which a cask can be lifted without additional 

safety analysis or design changes (e.g., addition of impact limiters) by the applicant. 

Reviewers should note that movement of cask system components within a reactor building 

may not meet the NRC’s criteria described in the NRC Bulletin 96-02, “Movement of Heavy 

Loads over Spent Fuel, over Fuel in the Reactor Core, or over Safety Related Equipment,” for 

movement of heavy loads within the reactor building. As such, if a potential user (licensee) has 

been identified, coordination with the appropriate project manager or technical lead from the 

NRC’s Office of Nuclear Reactor Regulation (NRR) should occur during the early stages of DSS 

design review. 

At a minimum, the NRC staff has generally addressed the conditions discussed below; however, 

other conditions may be relevant depending on specific details of the DSS design. Reviewers 

should pay particular attention to special design features and how these might be affected both 

by other external conditions and other DSS components. Reviewers should ensure all required 

information is provided in the SAR for the design earthquake accident analysis. 

“Normal” conditions (including conditions involving handling and transfer) and the extreme 

ranges of normal conditions are presumed to exist during design-basis accidents or designbasis natural phenomena with the exception of irrational or readily avoidable combinations. For 

example, an earthquake or tornado may occur at any time and in combination with any “normal” 

condition. By contrast, it can be presumed that transfer, loading, and unloading operations 

would not be conducted during a flood. 

“Off-normal” conditions and events are presumed to occur in combination with normal conditions 

that are not mutually exclusive. Nonetheless, it is not required that the SAR analyze or the 

system be designed for the simultaneous occurrence of independent off-normal conditions or 

events, design-basis accidents, or design-basis natural phenomena. 

Conditions involving a “latent” equipment or instrument failure or malfunction (that is, one that 

occurs and remains undetected) should be presumed to exist concurrently with other off-normal 

or design-basis conditions and events. Typical latent malfunctions include a misreading 

instrument that is not detected as part of routine procedures, an undetected ventilation 
 2-10 

blockage, or undetected damage from an earlier design-basis event or condition if no provisions 

exist for detection, recovery, or remediation of such conditions. 

For normal, off-normal, and accident-level conditions, reviewers should verify that the applicant 

has defined appropriate operating and accident scenarios. For these scenarios, the reviewer 

should verify the applicant includes in the SAR a comprehensive evaluation of the effects of 

such scenarios on the SSCs important to safety. The analyses of such events are addressed in 

individual chapters of the SRP. For example, the analyses of an earthquake on the DSS 

structural components are addressed in SRP Chapter 3, “Structural Evaluation.” The 

applicant’s evaluations should demonstrate that the requirements of 10 CFR §§ 72.104 and 

72.106 as well as 10 CFR Part 20 have been met. 

If appropriate, the following design bases should be included as operating controls and limits in 

SAR Chapter 13, “Technical Specifications and Operational Controls and Limits Evaluation”: 

 (1) Normal Conditions

For a given SNF specification, the primary external conditions that affect DSS 

performance are the ambient temperatures, insolation, and the operational 

environment experienced by the DSS. 

The NRC accepts as the maximum and minimum “normal” temperatures the 

highest and lowest ambient temperatures recorded in each year, averaged over 

the years of record. For the SAR, the applicant may select any design-basis 

temperatures as long as the restrictions they impose are acceptable to both the 

applicant and the NRC. If the cask is also designed for transportation, the 

temperature requirements of 10 CFR Part 71 could determine the design-basis 

temperatures for storage. 

For storage casks, the NRC staff accepts a treatment of insolation similar to that 

prescribed in 10 CFR Part 71.71 for transportation casks. If the applicant selects 

another design approach, the alternative approach should be justified in the SAR. 

The operational environment experienced by the DSS under normal conditions 

includes the manner in which the cask is loaded, unloaded, and lifted. 

Occupational dose rates will, in part, depend on whether the cask is sealed in a 

wet or a dry environment. Fuel cladding temperatures may also be affected. 

The manner in which the cask is lifted will determine the load on the trunnions 

and/or lifting yoke. The orientation of the cask (vertical or horizontal) and its 

height above ground during transport to the ISFSI will establish initial conditions 

for the drop accidents discussed below. 

 (2) Off-Normal Conditions

An applicant’s SAR generally addresses several off-normal conditions. These 

should include variations in temperatures beyond normal, failure of 10 percent of 

the fuel rods combined with off-normal temperatures, failure of one of the 

confinement boundaries, partial blockage of air vents, human error, out-oftolerance equipment performance, equipment failure, and instrumentation failure 

or faulty calibration. 
 2-11 

 (3) Accident Conditions

The staff has generally considered that the following accidents should be 

evaluated in the SAR. These do not constitute the only accidents that should be 

addressed if the SAR is to serve as a reference for accidents for a specific 

application. Other credible accidents that may be derived from a hazard analysis 

could include accidents resulting from operational error, instrument failure, 

lightning, and other occurrences. Post-accident recovery of damaged fuel may 

require such systems as overpacks or dry- transfer systems since ready retrieval 

of the fuel is required only for normal and off-normal conditions. Accident 

situations that are not credible because of design features or other reasons 

should be identified and justified in the SAR. Chapter 3, “Structural Evaluation” 

of this SRP provides more detail regarding accidents. 

 (a) Cask Drop 

The SAR should identify the operating environment experienced by the 

cask as well as the drop events (i.e., end, side, corner) that could result. 

Generally, the design basis is established either in terms of the maximum 

height to which the cask may be lifted when handled outside the reactor 

site SNF building or in terms of the maximum acceleration that the cask 

could experience in a drop. 

 (b) Cask Tipover 

Although cask system supporting structures may be identified and 

constructed as being important to safety (i.e., designed to preclude cask 

tipovers), the NRC considers that cask tipover events should be 

analyzed. In some cases, cask tipover may be determined to be a 

credible hazard, and the associated analysis should reflect the conditions 

(e.g., heights and accelerations) associated with that hazard. 

The NRC staff has accepted an unyielding surface for determining the 

bounding cask deceleration loads. Prototype or scale model testing and 

analytical modeling can be used. In the analytical approach, the hard 

receiving surface need not be unyielding. 

 (c) Fire 

The fire conditions postulated in the SAR should provide an “envelope” 

for subsequent comparison with site-specific conditions. The NRC 

accepts the methods discussed in 10 CFR 71.73(c)(4). In addition, the 

NRC staff accepts that the applicant may consider a fire based upon the 

limited availability of flammable material at an ISFSI (e.g., only that 

associated with vehicles transporting or lifting the cask, or sources of 

nearby combustible materials). Regardless of which approach the 

applicant takes, the SAR should specify and justify the bounding 

conditions for a “design-basis” fire. 
 2-12 

 (d) Fuel Rod Rupture 

The regulations require that the cask be designed to withstand the effects 

of accident conditions and natural phenomena events without impairing 

its capability to perform safety functions. Consequently, during the cask 

analysis for conditions resulting from design-basis accidents and natural 

phenomena, the NRC has asserted and the applicant should assume a 

release of 100 percent of the initial rod fill gases and a release of 

30 percent of the fission product gases from the fuel rods into the cask 

interior. The remaining 70 percent of the fission product gases is 

presumed to be retained within the fuel pellet. 

 (e) Leakage of the Confinement Boundary 

Casks are designed to provide the confinement safety function under all 

credible conditions. 

 (f) Explosive Overpressure 

The conditions under which a DSS may be exposed to the effects of an 

explosion vary greatly among individual sites. Generally, explosive 

overpressure is postulated to originate from an industrial accident. The 

NRC separately evaluated the effects of various sabotage methods on 

cask systems in developing appropriate regulations in 10 CFR Part 73. 

Consequently, this SRP does not consider explosive overpressures from 

sabotage events. 

The extent to which explosive overpressure is addressed in the SAR 

directly affects the degree of site-specific review required. The principal 

concern in the SAR should be the effects of explosive overpressure on 

the storage system rather than descriptions of hypothesized causes. 

Design parameters for blast or explosive overpressures should identify 

pressure levels as reflected (“side-on”) overpressure and provide an 

assumed pulse length and shape. This discussion should provide 

sufficient information for licensees to determine if the effects of their sitespecific hazards are bounded by the cask system design bases. 

 (g) Air Flow Blockage 

For storage systems with internal air flow passages, the reviewer should 

verify the applicant considers blockage of air inlets and outlets in an 

accident condition. The NRC staff considers that the effects of such an 

assumption should be utilized in determining the appropriate inspection 

intervals, and/or monitoring systems, for the DSS. 
 2-13 

 (4) Natural Phenomena Events (LOW Priority) 

The NRC staff has generally considered that the following events should be 

evaluated as design-basis accidents in the SAR: 

 (a) Flood 

The SAR should establish a design-basis flood condition. This condition 

may be determined on the basis of the presumption that the cask cannot 

tip over and the yield strength of the cask will not be exceeded. 

Alternatively, the SAR can show that credible flooding conditions have 

negligible impact on the cask design. 

If the SAR establishes parameters for a design-basis flood, all of the 

potential effects of flood water and ravine flood byproducts should be 

recognized. Serious flood consequences can involve effects such as 

blockage of ventilation ports by water and silting of air passages. Other 

potential effects include scouring below foundations and severe 

temperature gradients resulting from rapid cooling from immersion. 

 (b) Tornado 

The NRC staff accepts design-basis tornado wind loading as defined by 

RG 1.76, “Design Basis Tornado and Tornado Missiles for Nuclear 

Power Plants” (Region 1) and RG 1.117, “Tornado Design Classification.” 

Design criteria should be established for the cask on the basis of these 

wind-loading and missile-impact definitions. The cask should not tip over, 

and the capability to perform the confinement safety function should not 

be impaired. The NRC staff considers that tornados and tornado missiles 

may occur without warning. The review should note that, in general, the 

effects of a tornado missile bound those of a light general aviation aircraft 

directly impacting a DSS. 

 (c) Earthquake 

The SAR should state the parameters of the design earthquake. For use 

of a DSS at reactor sites, this is equivalent to the SSE used for analysis 

of nuclear facilities under 10 CFR Part 50. An analysis for an OperatingBasis Earthquake (OBE) is not required for a DSS SAR prepared in 

accordance with 10 CFR Part 72. Cask tipover accidents are analyzed, 

but tipover caused by an earthquake may not be a credible event. The 

reviewer should verify that the SSCs meet appropriate guidance in RG 

1.29, “Seismic Design Classification,” RG 1.61, “Damping Values for 

Seismic Design of Nuclear Power Plants,” and RG 1.92, “Combining 

Modal Responses and Spatial Components in Seismic Response 

Analysis.” 

 (d) Burial Under Debris 

Debris resulting from natural phenomena or accidents that may affect 

cask system performance may be addressed in the SAR or left to the site-
 2-14 

specific application. Such debris can result from floods, wind storms, or 

land slides. The principal effect is typically on thermal performance. 

 (e) Lightning 

Lightning typically has a negligible effect on cask systems; however, the 

requirements of the Lightning Protection Code and National Electric Code 

should be applied to the design of the cask system structures. The 

applicant should cite these codes as part of the general design criteria for 

the cask system (see Section 2.4.3.1). In addition, the SAR should 

address lightning as a natural phenomenon if cask-system performance 

may be impacted by the effect of lightning on a component that is 

important to safety. 

 (f) Other 

10 CFR Part 72 identifies several other natural phenomena events 

(including seiche, tsunami, and hurricane) that should be addressed for 

SNF storage. The SAR may include these natural phenomena as designbasis events or show that their effects are bounded by other events. If 

these events are not addressed in the SAR and they prove to be 

applicable to a specific site, a safety analysis is required prior to approval 

for use of the DSS under either a site-specific or general license. 

2.5.3 Design Criteria for Safety Protection Systems (MEDIUM Priority) 

Cask system components that are to be used in facility areas subject to review under 10 CFR 

Part 50 should satisfy both the requirements in 10 CFR Part 72 (with review guided by this SRP) 

and 10 CFR Part 50 (with review guided by NUREG-0800). Acceptance of the cask system in 

areas covered by 10 CFR Part 50 license requirements is not addressed in this SRP for 

approval under 10 CFR Part 72. If the applicant states that the storage system will be used at a 

specific reactor site, then the Division of Spent Fuel Storage and Transportation (SFST) project 

manager should inform the appropriate NRR project manager. The reviewer is reminded that 

heavy loads are a likely matter of interest to NRR. 

Table 2-2 presents a summary of design criteria (and design bases) that should generally be 

identified during the initial stages of the review. The applicability of Table 2-2 may vary 

depending on the details of the storage system design. 

Regardless of where the descriptions and associated criteria are located in the SAR, reviewers 

should include a description and evaluation of the safety protection systems in SER Chapter 2, 

“Principal Design Criteria.” The system descriptions should address the functions of the various 

system components in providing confinement, cooling, subcriticality, radiation protection of the 

public and workers, and SNF retrievability. Summary criteria for the performance of the system 

as a whole in providing for these capabilities or functions should also be described and 

evaluated. Reviewers should verify that the design-basis assumptions presented are consistent 

with and reasonable for actual site or facility conditions. Reviewers should also include a 

description and evaluation of the cask system design’s compatibility with removal from a reactor 

site, transportation, and ultimate disposition of the stored spent fuel. 
 2-15 

Table 2-2 Outline of Design Criteria and Bases for DSS 

Design Life • Limited to the requested term in the application 

Design Bases • SNF Specifications 

(1) Type 

(2) Configuration/Vendor 

(3) Enrichment (Maximum and Minimum) 

(4) Weight or Range of Weights 

(5) Burnup 

(6) Type of Cladding 

(7) Assemblies/Cask 

(8) Dimensions 

• Decay Heat/Assembly 

(1) Minimum Decay/Cooling Time (e.g., 5 years, 10 years, 

etc.) 

(2) Maximum Kilowatts per assembly 

• Gas Volume (at Temperature) 

• Fuel Condition/Damage Allowed 

• Burnup Credit 

(1) Credit Amount (specific actinides) 

(2) Operational History Parameters 

• Non-Fuel Hardware 

Normal Design Event 

Conditions

• Ambient Temperature 

(1) Maximum 

(2) Minimum 

• Loading 

(1) (Wet/Dry) 

• Storage Handling Orientation 

(1) (Vertical/Horizontal) 

• Maximum Lift Height 

• Maximum Cladding Temperature 

• Other Conditions Considered in 2.5.2.2 (1) 

Off-Normal Design 

Event Conditions

• Summarize Events Considered in 2.5.2.2 (2) 

Design-Basis 

Accident Design 

Events and 

Conditions

• End Drop 

(1) Lift Height (or Maximum Acceleration) 

• Side Drop 

(1) Lift Height (or Maximum Acceleration) 

• Tip-Over 

(1) Acceleration (if applicable) 

• Fire 

(1) Duration 

(2) Temperature 

• Other Events Considered in 2.5.2.2 (3) 

(As Applicable) 
 2-16 

Table 2-2 Outline of Design Criteria and Bases for DSS 

Design-Basis Natural 

Phenomena Design 

Events and 

Conditions

• Flood 

• Earthquake 

• Tornado 

• Other Events Considered in 2.5.2.2 (4) 

(As Applicable) 

 

Structural • Design Code (e.g., ASME, AISC) 

(1) Containment 

(2) Noncontainment 

(3) Basket 

(4) Trunnions 

(5) Storage Radiation and Protective Shielding and 

Enclosure 

(6) Transfer Radiation and Protective Shielding and 

Enclosure 

(7) Cooling Structure or System 

• Design Weight 

• Design Cavity Pressure 

(1) Normal/Off-Normal/Accident 

• Response and Degradation Limits 

(1) Normal/Off-Normal/Accident 

Thermal • Maximum Design Temperatures 

(1) Cladding 

(2) Other Components 

• Insolation (Side/Top/Bottom) 

• Fill Gas 

(1) Type (e.g., helium, etc.) 

(2) Initial Fill Pressure (at temperature) 

• Modes of Heat Transfer Utilized in the Design 

Confinement • Description of Confinement Boundary 

• Redundant Seals for Closure 

• Maximum Leak Rate for Confinement Boundary 

(1) Normal/Off-Normal/Accident 

(2) Justification of Leakage Rate if not Leaktight 

• Monitoring System Specifications 
 2-17 

Table 2-2 Outline of Design Criteria and Bases for DSS 

Radiation 

Protection/Shielding

• Confinement Cask 

(1) Surface Position 

Normal/Off-Normal/Accident 

• Exterior of Shielding 

(1) Transfer Mode Position 

(2) Storage Mode Position 

Normal/Off-Normal/Accident 

• ISFSI Controlled Area Boundary 

(1) Dose Rate 

(2) Annual Dose 

Normal/Off-Normal/Accident 

Criticality • Method of Control 

Geometry, Fixed Poison, Soluble Poison 

• Minimum Boron Concentration (Fixed and/or Soluble Poison) 

• Maximum keff

• Burnable Neutron Absorber Credit 

• Burnup Credit Analysis 

Materials • Cladding Hoop Stress 

• Corrosion 

Operating Procedures • Normal and Off-Normal 

• After Accident-level Conditions 

Acceptance Tests and 

Maintenance

• Industry codes and standards 

Tech Specs • Operational Controls and Limits 

Criteria relating to redundancy and allowable levels of response by the DSS under normal, offnormal, and accident-level conditions and events should be described and evaluated. In 

general, no unacceptable degradation in physical condition or functional performance should 

result from normal or off-normal conditions. The design criteria regarding limits of permissible 

system response and degradation resulting from an accident condition should be evaluated 

against the SSC capabilities to perform the principal safety functions. Considerations of 

permissible responses should include detect-ability and corrective actions that may be proposed 

as conditions of system use. 

The staff accepts that both routine surveillance programs and active instrumentation meet the 

intent of “continuous monitoring” as required in 10 CFR 72.122(h)(4). 

Reviewers should note that some DSS designs may contain a component or feature whose 

continued performance over the licensing period has not been demonstrated to staff with a 

sufficient level of confidence (e.g., rubber “O” rings). Therefore, staff may require the use of 

active instrumentation if the failure of that system or component causes an immediate threat to 

the public health and safety, and if that failure would not be detected by any other means. In 

some cases, to demonstrate compliance with 10 CFR 72.122(h)(4), the vendor or NRC staff 
 2-18 

may propose a technical specification requiring such instrumentation as part of the first use of a 

cask system. After first use, and if warranted and approved by staff, such instrumentation may 

be discontinued or modified. 

The staff should verify that the applicant has met the intent of continuous monitoring so that the 

applicant can determine when corrective action needs to be taken to maintain safe storage 

conditions. 

2.6 Evaluation Findings 

The reviewer will prepare evaluation findings on satisfaction of the regulatory requirements in 

Section 2.3. If the documentation submitted with the application supports positive findings for 

each of the regulatory requirements (the finding number is for convenience in reference within 

the SRP and SER), these statements should be similar to the following examples: 

 F2.1 The SAR and docketed materials adequately identify and characterize the SNF 

to be stored in the DSS in conformance with the requirements given in 

10 CFR 72.236. 

 F2.2 The SAR and the docketed materials relating to the design bases and criteria 

meet the general requirements as given in 10 CFR 72.122(a), (b), (c), (f), (h)(1), 

(h)(4), (i), and (l). 

 F2.3 The SAR and docketed materials relating to the design bases and criteria for 

structures categorized as important to safety meet the requirements given in 

10 CFR 72.122(a), (b)(1), (b)(2) and (b)(3), (c), (f), (h)(1), (h)(4), and (i); and 10 

CFR 72.236. 

 F2.4 The SAR and docketed materials meet the regulatory requirements for design 

bases and criteria for thermal consideration as given in 10 CFR 72.122 (a), 

(b)(1), (b)(2) and (b)(3), (c), (f), (h)(1), (h)(4), and (i). 

 F2.5 The SAR and docketed materials relating to the design bases and criteria for 

shielding, confinement, radiation protection, and ALARA considerations meet the 

regulatory requirements as given in 10 CFR 72.104(a) and (b); 10 CFR 

72.106(b); 10 CFR 72.122(a), (b), (c), (f), (h)(1), (h)(4), and (i); 10 CFR 

72.126(a). 

 F2.6 The SAR and docketed materials relating to the design bases and criteria for 

criticality safety meet the regulatory requirements as given in 10 CFR 72.124(a) 

and (b). 

 F2.7 The SAR and docketed materials relating to the design bases and criteria for 

retrievability meet the regulatory requirements as given in 10 CFR 72.122(a), 

(b)(1), (b)(2), and (b)(3), (c), (f), (h)(1), (h)(4), and (l). 

 F2.8 The SAR and docketed materials relating to the design bases and criteria for 

other SSCs not important to safety but subject to NRC approval meet the general 

regulatory requirements as given in the following subparts of 
 2-19 

10 CFR Part 72: Subpart E, “Siting Evaluation Factors” 72.104 and 72.106; 

Subpart F, “General Design Criteria” 72.122, 72.124, and 72.126; and Subpart L, 

“Approval of Spent Fuel Storage Casks.” 

The reviewer should provide a summary statement similar to the following: 

“The staff concludes that the principal design criteria for the [cask designation] are 

acceptable with regard to meeting the regulatory requirements of 10 CFR Part 72. This 

finding is reached on the basis of a review that considered the regulation itself, 

appropriate regulatory guides, applicable codes and standards, and accepted 

engineering practices. A more detailed evaluation of the design criteria and an 

assessment of compliance with those criteria are presented in Chapters 3 through 14 of 

the SER.” 
 3-1 

3 STRUCTURAL EVALUATION 

3.1 Review Objective 

In this portion of the dry storage system (DSS) review, the U.S. Nuclear Regulatory Commission 

(NRC) evaluates aspects of the DSS design and analysis related to structural performance 

under normal and off-normal operations, accident conditions, and natural phenomena events. 

In conducting this evaluation, the NRC staff seeks a high degree of assurance that the cask 

system will maintain confinement, subcriticality, radiation shielding, and retrievability or recovery 

of the fuel, as applicable, under all credible loads for normal and off-normal conditions 

accidents, and natural phenomenon events. 

3.2 Areas of Review 

This chapter of the DSS Standard Review Plan (SRP) provides guidance for use in evaluating 

the design and analysis of the proposed cask system with regard to its structural performance. 

All DSSs include a confinement cask that may have both internal components and integral 

external components. In addition, some DSSs have a variety of other components that are 

subject to NRC evaluation and approval under the cask certification provisions of Subpart L of 

10 CFR Part 72. 

Recognizing the diversity of the various cask system components, the NRC has broadly 

categorized the applicable review procedures and acceptance criteria as follows: 

 • Structural Capability of the Confinement boundary and Internals, 

 • Other structural system components important to safety, 

 • Other structural components subject to NRC approval. 

Within these broad categories, the NRC focuses the DSS structural evaluation, as described in 

Section 3.5, “Review Procedures,” using the following areas of review as appropriate: 

Scope

Structural Design Criteria and Design Features

 Design Criteria 

 General Structural Requirements 

 Applicable Codes and Standards 

 Structural Design Features 

Materials Related to Structural Evaluation 

 Structural Analysis

 Load Conditions 

 Normal Conditions 

 Off-normal Conditions 

 Natural Phenomena and Accident Conditions 

 Structural Analysis Methods 

 Finite-element Analysis 

 Closed-form Calculations 

Structural Analysis for Specific Cask Components 

 Structural Evaluation 
 3-2 

 Structural Capability 

 Fabrication and Construction 

3.3 Regulatory Requirements 

Table 3-1 presents a matrix that shows the primary relationship of the regulations provided in 

this section to the specific areas of review associated with this SRP chapter. The NRC staff 

reviewer should verify the association of regulatory requirements with the areas of review 

presented in the matrix to ensure that no requirements are overlooked as a result of unique 

applicant design features. 

Table 3-1 Relationship of Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Regulations

72.124(a) 72.234(a), (b) 72.236(b),(c), 

(d), (l) 72.236(g), (h) 

Scope ● ● ● 

Structural Design Criteria 

and Design Features ● ● ● ●

Materials Related to 

Structural Evaluation ● 

Structural Analysis ● ● 

Structural Evaluation ● ● ●

3.4 Acceptance Criteria 

The most important function of the structural analysis is to ensure sufficient structural capability 

for every applicable section of the cask system to withstand the worst-case loads under 

accident conditions and natural phenomena events. Withstanding such loads enables the cask 

system to successfully preclude the following negative consequences: 

 • Unacceptable risk of criticality, 

 • Unacceptable release of radioactive materials, 

 • Unacceptable radiation levels, 

 • Impairment of retrievability or recovery, as applicable. 

Because of the diversity of cask system components and various materials that are subject to 

NRC evaluation and approval, it is not possible to define objective structural review criteria that 

address all possible component configurations. No single structural code currently accepted by 

the NRC (such as the American Society of Mechanical Engineers [ASME] Boiler and Pressure 

Vessel [B&PV] Code, Section III, Division 1 [ASME B&PV]) or Section III, Division 2 may cover 

the design of all spent nuclear fuel (SNF) storage systems. Consequently, the acceptability of 

any given structure will be contingent upon a combination of adherence to applicable portions of 

multiple codes and a review of the functional performance of the structure taken as a whole. 
 3-3 

This combined approach allows the designer to request relief, or provide alternatives, and the 

reviewer to impose additional restrictions when warranted by specific design features. 

In general, the DSS structural evaluation seeks to ensure that the proposed design and analysis 

fulfill the following acceptance criteria that reflect the industry codes and standards the NRC 

staff has accepted in past DSS structural evaluations. The American National Standards 

Institute’s (ANSI) “Design Criteria for an Independent Spent Fuel Storage Installation (Dry 

Storage Type)” (ANSI/ANS-57.9) generally applies to the design and construction of an ISFSI 

but contains some criteria/design requirements relative to dry storage systems. 

3.4.1 Confinement Cask and Metallic Internals 

3.4.1.1 Steel Confinement Cask 

The structural design, fabrication, and testing of the confinement system and its redundant 

sealing system should comply with an acceptable code or standard such as ASME B&PV Code. 

(The NRC has accepted use of either Subsection NB or Subsection NC of Section III, Division 1 

of this code.) Division 3 of Section III of the ASME B&PV Code, addressing storage of spent 

nuclear fuel, has been published, but currently no NRC position has been established on that 

standard. Other design codes or standards may be acceptable depending on their application. 

An applicant must justify the use of new criteria where no NRC staff position has been 

established. 

 i. The NRC staff evaluates the proposed limitations on allowable stresses and 

strains in the confinement cask, steel parts important to safety and subject to 

review by comparison with those specified in applicable codes and standards. 

Where certain proposed load combinations will produce values that exceed the 

accepted limits for localized points on the structure, the applicant should provide 

adequate justification to show that the deviation will not affect the functional 

integrity of the structure. Under certain conditions limiting strains and limiting 

deformations may form part of the acceptance criteria. 

 ii. The NRC has accepted the use of applicable subsections of the ASME B&PV 

Code, Section III, Division 1, such as Subsections NF and NG, for components 

used in the cask system. This includes the “basket” structure used in casks to 

restrain and position multiple fuel elements in the storage system in which 

Subsection NG has been used. 

3.4.1.2 Steel-Lined Concrete Confinement Cask 

 i. The American Concrete Institute (ACI) and ASME’s “Code for Concrete Reactor 

Vessels and Containments” (ACI 359), also known as Section III, Division 2 of 

the ASME B&PV Code, constitutes an acceptable standard for prestressed and 

reinforced concrete structures that are an integral component of a steel-lined 

concrete confinement cask that must withstand internal pressure in operation or 

testing and constitutes a confinement cask. The minimum functional 

requirements of ANSI/ANS-57.9 for subject areas not specifically addressed in 

ACI 359 shall be met. 

ii. The NRC will review the use of applicable subsections of the ASME B&PV Code, 

Section III, Division 1, such as Subsections NF and NG, for components used 
 3-4 

within the confinement cask but not integrated with it. This includes Subsection 

NG for the “basket” structure used in casks to restrain and position multiple fuel 

elements in the storage system. 

3.4.2 Other Structural System Components and Structures Important to Safety 

The NRC accepts the use of ANSI/ANS-57.9 (together with the codes and standards cited 

therein) as the basic reference for the ISFSI dry storage systems important to safety that are not 

designed in accordance with accepted provisions or alternatives to applicable portions of 

Section III, Division 1 or 2 (ACI-359) of the ASME B&PV Code. Structures and components that 

are important to safety which are related to lifting and handling cask systems should comply 

with American National Standards Institute (ANSI) Standard, “American National Standards for 

Radioactive Material Lifting Devices for Shipping Containers Weighing 10,000 lbs (4500 kg) or 

More” (N14.6). The loadings defined in American Society of Civil Engineers, “Minimum Design 

Loads for Buildings and Other Structures,” (ASCE 7) can be used when load combinations are 

considered on the basis of ANSI/ANS-57.9. 

3.4.2.1 Steel Structures 

The principal codes and standards include the following references that may be applied to steel 

structures and components: 

a. American Institute of Steel Construction (AISC), “Specification for Structural Steel 

Buildings — Allowable Stress Design and Plastic Design.” 

b. AISC, “Load and Resistance Factor Design Specification for Structural Steel 

Buildings.” 

c. American Welding Society, “Structural Welding Code Steel,” (AWS D1.1). 

3.4.2.2 Reinforced Concrete Structures 

ACI’s “Code of Requirements for Nuclear Safety Related Concrete Structures,” ACI 349 can be 

applied to reinforced concrete structures and components. 

3.4.3 Other Structural Components Subject to NRC Approval 

For structural design and construction of other components subject to NRC approval, the 

principal codes and standards include the following: 

a. American Society of Civil Engineers (ASCE), “Minimum Design Loads for 

Buildings and Other Structures” (ASCE 7). 

b. International Building Code (IBC) 2006 from International Code Council. 

c. AISC, “Specification for Structural Steel Buildings—Allowable Stress Design and 

Plastic Design.” 

d. AISC, “Code of Standard Practice for Steel Buildings and Bridges.” 

e. ASME B&PV Code, Section VIII. 
 3-5 

 f. ACI 318, “Building Code Requirements for Structural Concrete.” 

3.5 Review Procedures (HIGH Priority) 

The SAR documentation should be reviewed to confirm that the design of the cask structure 

provides for satisfactory functional performance. This includes operating suitability within 

specified limiting conditions and satisfaction of the basic safety criteria under all credible events 

and environmental conditions. 

The SAR should clearly identify the confinement system and other structures important to 

safety, and each component should have sufficient structural capability for every applicable 

section to withstand the worst-case loads under accident-level events and conditions to 

successfully preclude the following: 

• Unacceptable risk of criticality. 

• Unacceptable release of radioactive materials to the environment. 

• Unacceptable radiation dose to the public or workers. 

• Significant impairment of retrievability or recovery, as applicable, of stored 

nuclear materials (the NRC has accepted some degradation of retrievability 

under accident conditions and severe natural phenomena events that are treated 

as design bases events). 

This position does not necessarily require that all confinement system and other structures 

important to safety survive all design-basis accidents and extreme natural phenomena without 

any permanent deformation or other damage. Some load combination expressions for the 

design basis event (DBE) and conditions for structures important to safety permit stress levels 

that exceed yield. The SAR should include computations of the maximum extent of potentially 

significant accident deformations and any permanent deformations, degradation, or other 

damage that may occur. The reviewer should verify that the applicant has performed 

computations, analyses, and/or tests and that both the tests and results are acceptable to the 

NRC to clearly demonstrate that any permanent deformations, degradation, or other damage 

that may occur does not render the system performance unacceptable. 

Structures important to safety are not required to survive accidents to the extent that they 

remain suited for use for the life of the cask system without inspection, repair, or replacement. If 

the service life of structures important to safety may be degraded by accident-level conditions, 

there must be SAR commitments and procedures for determining and correcting the 

degradation and performing other acceptable remedial action. 

The proposed technical specifications should be reviewed to ensure that they include adequate 

restrictions on cask handling and operations to preclude the possibility of damage to the 

structure or the confined nuclear material. Operating controls and limits of the technical 

specifications (reviewed under Chapter 13 of this SRP) should be included in both the SAR and 

the SER, and should describe actions to be taken and inspections to be conducted upon 

occurrence of events that may cause such damage. 
 3-6 

Figure 3-1 presents an overview of the evaluation process and can be used as a guide to assist 

in coordinating with other review disciplines. 

In evaluating the structural design and performance of a proposed DSS, the reviewer should 

select and emphasize aspects of the following review procedures, as appropriate for the 

particular DSS, in relation to the acceptance criteria summarized in Section 3.4. 

Description of Structures, Systems, and Components Important to Safety

The reviewer should verify that the applicant’s safety analysis report (SAR) clearly identifies the 

proposed structural design and construction of structures, systems, and components (SSCs) 

that are important to safety and necessary for effective functional performance and safety of the 

DSS. The SAR and supplemental material submitted by the applicant should be reviewed to 

assess compliance with the applicable scope and content requirements defined in 10 CFR 

72.230. The reviewer should focus in particular on requirements and conditions of use related 

to design, construction, implementation, operation, and maintenance of structural SSCs. 

Applicable Codes, Standards, and Specifications

NRC guidelines recommend that the safety evaluation report (SER) prepared by the NRC staff 

include a table (in the design criteria evaluation section) summarizing the applicable reference 

sources. This table should identify all source documents cited in the SAR, their usage (e.g., 

description of model, prior NRC approval of cask system elements, design code, construction 

code), and acceptability for that usage. The sources of interest include documents directly 

referenced in the SAR; sources of material incorporated by reference; and codes, standards, 

specifications, and other sources of criteria that further define the design and construction of the 

proposed structures. If not tabulated, the consolidated review and assessment of reference 

sources should otherwise be included in the SER. 

Loads and Load Combinations

The reviewer should verify that the loads and load combinations are as specified in Chapter 2, 

“Principal Design Criteria Evaluation,” of this SRP. If the applicant has not adequately justified 

any deviations from the acceptance criteria for loads and load combinations, the reviewer 

should identify the deviations as unacceptable and transmit them to the applicant for further 

justification. If components associated with or integral to the fuel assembly are to be stored in 

the cask, then the reviewer should ensure these components are considered by the applicant in 

the structural analyses. 
 3-7 

Figure 3-1 Overview of the Structural Evaluation

Chapter 1 – General 

Information Evaluation 

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation 

• DSS Description and 

Operational Features 

• Drawings 

• DSS Contents 

• Codes and Standards 

• Codes and Standards 

• Structural and Pressure 

Tests 

• Component Tests 

Chapter 12 – Accident 

Analysis Evaluation 

• Cause of Event 

• Detection of Event 

• Summary of Event 

Consequences and 

Regulatory Compliance 

• Corrective Course of 

Action 

• Ready Retrievability 

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation 

• Operating Limits 

• Limiting Conditions 

• Design Features 

Chapter 6 – Shielding 

Evaluation 

• Design Criteria 

• Design Features 

• Material Properties 

Chapter 7 – Criticality 

Evaluation 

• Criticality Design 

Criteria and Features 

Chapter 9 – Operating 

Procedures Evaluation 

• Closure Configuration 

• Lifting Configuration 

• Bolt Torque 

• Tie-down Configuration 

• Handling Restrictions 

• Loading Configuration 

• Structural Evaluation 

– Structural Capability 

– Fabrication and 

Construction 

Structural Analysis

• Load Conditions 

– Normal Conditions 

– Off-normal Conditions 

– Accident Conditions and Natural 

Phenomena 

• Structural Analysis Methods 

– Finite-element Analysis 

– Closed-form Calculations 

– Structural Analysis for Specific Cask 

Components 

Scope 

Structural Design Criteria and 

Design Features 

• Design Criteria 

– General Structural 

Requirements 

– Applicable Codes and 

Standards 

• Structural Design Features 

Related to Structural Evaluation 

Chapter 3 – Structural Evaluation

Chapter 2 – Principal 

Design Criteria Evaluation 

• Spent Fuel Specifications 

• General Design Criteria 

• Structural Design Criteria 

• Thermal Design Criteria 

• Shielding/Confinement/ 

Radiation Protection 

Design Criteria 

• Criticality Design Criteria 

• Material Capability 

Chapter 4 – Thermal 

Evaluation 

• Temperature 

• Pressure 

Chapter 8 – Materials 

Evaluation 

• Applicable Codes and 

Standards 

• Material Properties 

• Weld Design and 

Specification 

• Bolt Applications 

• Coatings 

• Concrete and Reinforcing 

Steel 

• Seals 

• Loss of Corrosion 

Resistance 

Chapter 5 –

Confinement Evaluation 

• Design Features 

Chapter 8 – Materials 

Evaluation 

• Structural Materials 

• Structural Design 

Criteria 

• Structural Compatibility 

with Functional 

Performance 
 3-8 

The SAR should include a comprehensive table of load combinations and safety margins for 

selected structural sections of components important to safety (or otherwise subject to NRC 

evaluation). The summary table should include sufficient structural sections and forms of 

loadings (e.g., shear, flexure, axial, and combined stress situations) to verify that the lowest 

margins of safety are represented for the various components. In addition, this table can be 

used to summarize the structural capacity evaluation. 

Design and Analysis Procedures

The reviewer should determine whether the applicant’s design and analysis procedures and 

assumptions are conservatively defined on the basis of accepted engineering practice. The 

behavior of the structure under various loads, and the manner in which these loads are treated 

in conjunction with other coexistent loads should be reviewed, while compliance with the 

acceptance criteria, defined in Section 3.4 of this SRP should be assessed. 

Structural Acceptance Criteria

The proposed limitations on allowable stresses, strains, or deformations in the confinement 

cask, its internals, system components important to safety, and other components subject to 

review should be analyzed. The reviewer should compare the proposed limitations with those 

specified in the applicable codes and standards. Where the applicant proposes to exceed the 

accepted limits for certain load combinations at localized points on the structure, the reviewer 

should evaluate the justification provided to ensure that the deviation will not affect the 

functional integrity of the structure. If the justification is not acceptable, the reviewer should 

request additional justification and bases. 

Materials, Quality Control, and Special Fabrication Techniques

Information provided in the SAR regarding materials is reviewed under the guidance of Chapter 

8, “Materials Evaluation” of this SRP. Quality control methods, and special fabrication 

techniques, if any, related to the structural evaluation should be reviewed in coordination with 

the materials discipline and QA. The QA program is reviewed under Chapter 14 “Quality 

Assurance Evaluation” of this SRP. If the applicant proposes to use a new material not 

addressed in prior approvals, the applicant must provide sufficient data regarding the material’s 

structural properties to establish the acceptability of the material. Similarly, the reviewer should 

evaluate any new quality control programs or construction techniques to ensure that they will 

not degrade the structural quality, integrity, or function of the DSS. 

Testing and In-Service Surveillance Requirements

The proposed pressure test procedures for the confinement cask should be reviewed in 

comparison with the procedures described in ASME Code, Section III, Division 1, Subsection 

NB-6000, and in conjunction with Chapter 10, “Acceptance Tests and Maintenance Program 

Evaluation” of this SRP. Also, the proposed acceptance test and maintenance requirements for 

trunnions should be reviewed in comparison with those described in the ASME Code and ANSI 

N14.6, as applicable for load bearing components. Any other proposed testing and in-service 

surveillance programs should be reviewed on a case-by-case basis. Also, the reviewer should 

read SAR Section 10 to verify that the applicant has included all appropriate acceptance tests 

and addressed all required evaluations in Section 10 of the SER. 
 3-9 

Conditions for Use of Structures

The structural evaluation should be reviewed to determine if conditions of use or technical 

specifications should be associated with the structural design or proposed fabrication and 

construction methods. The reviewer should determine the appropriateness of and need for any 

proposed technical specifications related to structural design and construction. Also, the 

reviewer should determine whether any additional technical conditions related to structural 

performance are needed and, if so, provide input to the conditions of use discussed in the SER. 

In addition, the reviewer should describe the basis for the suggested conditions in the structural 

evaluation section of the SER. Structure-related conditions of use may be linked to evaluations 

performed under other sections (such as a field verification that maximum concrete 

temperatures predicted from thermal analysis will not be exceeded). 

The remainder of this section provides specific review procedures for each of the three 

categories of cask system components including the confinement cask and steel internals, other 

structures important to safety, and other components subject to NRC approval. Within each of 

these broad categories, the specific review procedures focus the DSS structural evaluation 

using the areas of review identified in Section 3.2 of this SRP. 

3.5.1 Confinement Cask and Metallic Internals 

The structural review of the confinement cask addresses drawings, plans, sections, supporting 

computations, and specifications for those structural components comprising confinement 

barriers. The review also addresses structural and sealing interfaces, and connections that are 

necessary to complete the confinement system (as defined in 10 CFR Part 72). In addition, this 

review includes evaluation of components that serve no structural function to confirm that they 

do not impair the functioning of the confinement cask. The review also encompasses the 

evaluation of the metallic internals that constitute the “basket” structure. 

3.5.1.1 Scope 

The SAR must describe all components of the confinement cask and internals important to 

safety in sufficient detail to allow evaluation of their structural behavior and effectiveness under 

the imposed design conditions. In addition, the SAR must identify all codes and standards 

applicable to the components. 

The discussion in the SAR must demonstrate that all components of the confinement cask and 

internals important to safety will be designed and fabricated to quality standards commensurate 

with the importance to safety of the function to be performed. In addition, components of the 

confinement cask and internals important to safety must be designed to accommodate the 

combined loads anticipated during normal, off-normal, accident, and natural phenomenon 

events with an adequate margin of safety. 

3.5.1.2 Structural Design Criteria and Design Features 

i. Design Criteria (MEDIUM Priority) 

The cask-related design criteria presented in SAR Chapter 2, “Principal Design 

Criteria Evaluation” should be reviewed as well as the criteria provided herein. 

The NRC generally considers the following design criteria to be acceptable to 

meet the structural requirements of 10 CFR Part 72: 
 3-10 

(1) General Structural Requirements 

The proposed cask must maintain confinement of radioactive material 

under normal and off-normal operations, accident conditions, and natural 

phenomenon events. In addition, neither the cask nor any basket within 

the cask may deform under credible loading conditions in a manner that 

would jeopardize the subcritical condition and recovery or retrievability of 

the fuel, as applicable. 

The design must adequately protect the fuel cladding against gross 

rupture caused by degradation resulting from design or accident 

conditions. In addition, the design must ensure that the SNF will not 

experience accelerations/decelerations that would damage its structural 

integrity or jeopardize its subcritical condition or retrievability under 

normal and off-normal design conditions. 

The applicant must analyze the cask to show that it will not tip over or 

drop in its storage condition as a result of a credible natural phenomenon 

event. A tipover or drop is always assessed as a bounding condition 

during handling operations. 

Radiation shielding in the cask system is required to protect the public 

and workers involved with spent fuel handling and storage, and such 

shielding must not degrade under normal or off-normal conditions or 

events. The shielding function may degrade as a result of an accident 

(e.g., displacement of source or shielding, reduction in shielding). 

However, the loss of function must be readily visible, apparent, or 

detectable. (Any permissible degradation in shielding must be shown to 

result in dose rates sufficiently low to permit recovery of the damaged 

cask including unloading, if necessary). The necessary structural criteria 

to assure adequate shielding remains in-place should be clearly 

identified. 

(2) Applicable Codes and Standards 

The structural design, fabrication, and testing of the confinement system 

and any necessary redundant sealing system should comply with 

acceptable codes or standards. Use of codes and standards previously 

accepted by the NRC expedites the evaluation process. Use of other 

codes and standards, definition of criteria composed of extracts from 

multiple codes and standards with overlapping scopes, or substitution of 

other criteria, in whole or in part, in place of acceptable published codes 

or standards requires a custom NRC review and may delay the evaluation 

process. 

Section III, Division 1, of the ASME B&PV Code is an accepted code for 

design, fabrication, and test of steel confinement casks. Specifically, the 

NRC has accepted use of either Subsection NB or NC. Other design 

codes or standards may be acceptable depending on their application. 

The NRC has accepted use of the applicable subsections of the ASME 
 3-11 

Code, Section III, Division 1, for cask system components used within the 

confinement cask but not integrated with it. This includes the “basket,” 

which is a structure used in casks to restrain and position multiple fuel 

elements. Section III, Division 3 of the ASME B&PV Code is also 

available and addresses storage cask systems, but NRC has not 

endorsed its use at the current time. 

Also, the NRC has accepted applicable subsections of Division 1, of the 

ASME Code, for structural external integral elements of the confinement 

(e.g., Subsection NF for integral supports) cask. 

Commitments for structures important to safety to ASME Code Section III, 

with proposed alternatives to the Code, should be documented in the 

application. Likewise, NRC staff-approved alternatives to the Code 

should be incorporated, either directly or referenced, in the certificate of 

compliance (or in the technical specifications attached to the certificate) 

issued by the NRC. In the event that alternatives to codes are required 

during fabrication and the alternatives do not impact the quality or safety 

of the component, an alternative to the requirements of the certificate of 

compliance or technical specification may be granted with approval of the 

NRC. 

Applicants should propose a condition to the certificate of compliance or 

technical specification, either directly or referenced, describing the 

alternatives to the referenced codes. The condition or technical 

specification should also describe a process to address one-time 

alternatives from the ASME Code that may occur during fabrication. The 

information provided should include the identification of the component, 

the reference to the ASME Code (code edition, addenda, section or 

article), description of the Code requirement, and a description of the 

alternative. In addition, the applicant should justify the alternative, 

including a description of how the alternative would provide an acceptable 

level of quality and safety. Additionally, the applicant should describe 

how compliance with the code provisions would result in hardship or 

difficulty without a compensating increase in the level of quality or safety. 

For a steel-lined concrete confinement cask system, NRC accepts ACI 

359, also designated Section III, Division 2, of the ASME Boiler and 

Pressure Vessel Code. This Code is acceptable for prestressed and 

reinforced concrete that is an integral component of a radioactive material 

containment vessel that must withstand internal pressure in operation or 

testing. ACI 359, as endorsed by RG 1.136, Rev. 3, “Design Limits, 

Loading Combinations, Materials, Construction, and Testing of Concrete 

Containments,” and Section 3.8.1, “Concrete Containments” of NUREG0800, “Standard Review Plan for Review of Safety Analysis Reports for 

Nuclear Power Plants,” should be applied on the basis of containment 

function regardless of whether the concrete structure is fixed or portable 

and regardless of where the concrete structure is fabricated. ACI 359 

also applies to structural concrete supports constructed as an integral 

part of the containment. If ACI 359 and RG 1.136 apply to the structure, 
 3-12 

the Code applies to the entire design, material selection, fabrication, and 

construction of that structure. 

As an alternative to the requirements of Section CC-3440 of ACI 359, the 

NRC also accepts the following. These criteria are an alternative to the 

temperature requirements of ACI 349, A.4, but only for the specified uses 

and temperature ranges: 

a. If concrete temperatures of general or local areas are 93°C 

(200°F) in normal or off-normal conditions/ occurrences, no tests 

to prove capability for elevated temperatures or reduction of 

concrete strength are required. 

b. If concrete temperatures of general or local areas exceed 93°C 

(200°F) but would not exceed 149°C (300°F), no tests to prove 

capability for elevated temperatures or reduction of concrete 

strength are required if Type II cement is used and aggregates are 

selected which are acceptable for concrete in this temperature 

range. The following criteria for fine and coarse aggregates are 

acceptable: 

1) Satisfy ASTM C33, (“Standard Specification for Concrete 

Aggregates”) requirements and other requirements 

referenced in ACI 349 for aggregates. 

2) Satisfy ASTM C150, (“Standard Specification for Portland 

Cement”) requirements and other requirements referenced 

in ACI 349 for cement. 

3) Have demonstrated a coefficient of thermal expansion 

(tangent in temperature range of 20°C to 38°C [70°F to 

100°F]) no greater than 11x10-6 mm/mm/°C (6x10-6

in./in./°F), or be one of the following minerals: limestone, 

dolomite, marble, basalt, granite, gabbro, or rhyolite. 

c. If concrete temperatures of general or local areas in normal or offnormal conditions or occurrences do not exceed 107°C (225°F), 

the requirements of 1 and 2 apply to the coarse aggregate, but 

fine aggregate that meets 1 and is composed of quartz sands or 

sandstone sands may be used in place of compliance with 2. 

ii. Structural Design Features (LOW Priority) 

The cask-related descriptive information presented in SAR Chapter 1, “General 

Information Evaluation” should be reviewed as well as any related information 

provided in SAR Chapter 3 “Structural Evaluation”. The drawings, figures, tables, 

and specifications included in the SAR should fully define the structural features 

of the cask. These may include the cask system that could include an inner 

shell, an outer shell, and a gamma shield, inner and outer lids and bolts, port 

covers and bolts, vent port covers to be welded in place, neutron shields and 

shell, trunnions, fuel basket, and impact limiters (if used). 
 3-13 

The reviewer should coordinate with the confinement review (Chapter 5, 

“Confinement Evaluation,” of this SRP) to verify that the SAR clearly identifies the 

confinement boundaries. These boundaries include the primary confinement 

vessel; its penetrations, seals, welds, and closure devices; and the redundant 

sealing system as provided by the system. 

The list of weights and calculation of the cask center of gravity should be 

reviewed. The reviewer should verify that the applicant used the appropriate 

limiting cases in the structural evaluations. 

3.5.1.3 Materials Related to Structural Evaluation (HIGH Priority) 

The structural reviewer should coordinate with the materials reviewer to determine the impact of 

corrosion, reviewed in Chapter 8, “Materials Evaluation” of this SRP, on structural integrity. The 

reviewer should ensure that the applicant used appropriate corrosion allowances for the 

structural analyses. The reviewer should consider the static and dynamic (where appropriate) 

stresses, strains, deformations, and response, and the limits used for the structural design and 

evaluations. 

A DSS serves to confine and maintain safe storage conditions throughout its service life. 

Design and construction codes (e.g., ASME B&PV Code Section III) give reasonable assurance 

that the as-fabricated material will provide the necessary integrity. It is noted that the ASME 

Code Section III, Division 1, applies specifically to maintaining pressure boundaries and 

supporting structures in nuclear power plants, and may not necessarily be totally applicable to 

all DSS. However, designers may choose to cite it as the code to which selected components 

are to be fabricated. Codes such as the ASME B&PV are not likely to address all the potential 

performance conditions (e.g., cracking, creep, corrosion, etc.) that may arise from 

environmental, electrochemical, or dynamic-loading. These and other effects are specific to the 

individual application and should be addressed to meet the guidance of Chapter, 8, “Materials 

Evaluation” of this SRP. 

The reviewer should verify that the properties used are appropriate for the load conditions of 

interest (e.g., static or dynamic, impact loading, hot or cold temperature, wet or dry conditions). 

SAR Chapter 12, “Accident Analyses Evaluation” should be reviewed to ensure that the 

applicant considered any appropriate restrictions regarding temperature or environmental 

conditions for the materials under accident conditions. 

The reviewer should coordinate with the thermal and material disciplines to determine the 

appropriate temperatures at which allowable stress limits should be defined. For most cask 

materials, the stress limits should be defined at the maximum temperature for each material as 

established by the SAR thermal analysis. Further discussion of the background for the 

temperature limits can be found in Chapters 4, “Thermal Evaluation” and 8, “Materials 

Evaluation” of this SRP. 

The reviewer should coordinate with the materials, criticality, and shielding reviews to ensure 

that, during storage and accident conditions, any structural materials considered as neutron 

absorbers and/or gamma shields will perform safety functions as intended. 

If the cask has impact limiters, used in the transfer and storage operations, the applicant should 

thoroughly evaluate and verify their nonlinear impact characteristics. In addition, the applicant 
 3-14 

should tabulate and describe the crush characteristics and properties of the limiters in the 

directions that are to be used. 

3.5.1.4 Structural Analysis 

i. Load Conditions 

(MEDIUM Priority) To meet the structural requirements of 10 CFR Part 72, the 

DSS design must accommodate the full spectrum of load conditions including all 

anticipated normal, off-normal, and accident-level conditions (including natural 

phenomenon events). The system should not experience any permanent 

deformation or loss of safety function capability during normal or off-normal 

operating conditions. However, the system may experience some permanent 

deformation, but no loss of safety function capability, in response to an accident. 

(1) Normal Conditions (LOW Priority) 

Normal conditions and events are those associated with cask system 

operations, including storage of nuclear material, under the normal range 

of environments. The SAR should state the assumed limits of normal use 

environments to support evaluation by a user of the certified cask system 

suitability for use at a specific site under a general license. 

Loads normally applicable to a confinement cask include weight, internal 

and external pressures, and thermal loads associated with operating 

temperature. The loads experienced may vary during loading, 

preparation for storage, transfer, storage, and retrieval operations. The 

weight is the maximum or design weight (including tolerances) of the cask 

as it is stored and loaded with SNF. However, depending on the 

operation and procedures, the weight should also include water fill. The 

applicant should evaluate all orientations of the cask body and closure 

lids during normal operations and storage conditions including loads 

associated with loading, transfer, positioning, and retrieval of the 

confinement cask. 

Internal pressures result from hydrostatic pressure, cask drying and 

purging operations, filling with non-reactive cover gas, out-gassing of fuel, 

refilling with water, radiolysis, and temperature increases. Temperature 

variations and thermal gradients in the structural material may cause 

additional stresses in the cask and closure lids. The reviewer should 

coordinate with the thermal review (Chapter 4, “Thermal Evaluation,” of 

this SRP) to determine the conservative (or enveloping) values and 

combinations of the cask internal pressures and temperatures for both hot 

and cold conditions. The reviewer should use the temperature gradients 

calculated in SAR Chapter 4 to determine thermal stresses. Note that if 

the confinement system has several enclosed areas; all areas may not 

have the same internal pressures. In some casks, enclosed areas 

consist of the cask cavity and the region between the inner and outer lids. 

Required evaluations include weight plus internal pressures and thermal 

stresses from both hot and cold conditions. The reviewer should verify 
 3-15 

that the applicant included the maximum thermal gradient as determined 

in the thermal analysis, when evaluating thermal stresses. 

(2) Off-Normal Conditions (LOW Priority) 

The review should identify and evaluate all off-normal events and 

conditions described in Chapter 12, “Accident Analyses Evaluation,” of 

this SRP. The off-normal conditions and events should be reviewed for 

those that affect the confinement cask structure. The confinement cask 

components should satisfy the same structural criteria required for normal 

conditions, as discussed above. 

The SAR should clearly identify anticipated off-normal conditions and 

events that may reasonably be expected to occur during the life of the 

cask system at the proposed site. In addition, the SAR should state the 

environmental limits to support comparison of the cask system design 

bases with specific site environmental data. Off-normal conditions and 

events can involve potential mishandling, simple negligence of operators, 

equipment malfunction, loss of power, and severe weather (short of 

extreme natural phenomena). 

(3) Accident-Level Events and Conditions 

The reviewer should follow the guidance below in reviewing the structural 

response to accident-level conditions. Note that the SAR must address, 

at a minimum, each of the following accidents. However, this discussion 

may not address all of the potential events or accidents that apply to a 

cask (Chapter 12 of this SRP addresses the identification and evaluation 

of accidents). 

(a) Cask Drop and Tipover (HIGH Priority) 

The reviewer should ensure the applicant performs a cask drop 

and tipover analysis or demonstrates that this scenario is not 

credible. The SAR should identify the operating environment 

experienced by the cask and the drop events (end/side/tipover) 

that could result. Generally, applicants establish the design basis 

in terms of the maximum height to which the cask is lifted outside 

the building or the maximum deceleration that the cask could 

experience in a drop. The design-basis drops should be 

determined on the basis of the actual potential handling and 

transfer accidents. 

If the analytical approach described in the LLNL report 

UCID-21246 (Chun, R., et al., 1986) for axial buckling is used to 

assess fuel integrity for the cask drop accident, the analysis 

should use the irradiated material properties and should include 

the weight of fuel pellets. 

Alternatively, an analysis of fuel integrity which considers the 

dynamic nature of the drop accident and any restraints on fuel 
 3-16 

movement resulting from cask design is acceptable if it 

demonstrates that the cladding stress remains below yield. If a 

finite element analysis is performed, the analysis model may 

consider the entire fuel rod length with intermediate supports at 

each grid support (spacer). Irradiated material properties and 

weight of fuel pellets should be included in the analysis. 

The NRC will accept cask tipover about a lower corner onto a hard 

receiving surface from a position of balance with no initial velocity. 

The NRC has also accepted analysis of cask drops with the 

longitudinal axis horizontal which, together with analysis of a 

vertical drop, could bound a non-mechanistic tipover case. 

NRC staff has accepted an unyielding surface for determining the 

bounding cask deceleration loads that can far exceed the 

decelerations experienced by a cask dropping onto or tipping over 

the concrete storage pad that will bend and deform. Prototype or 

scale model testing can be used to obtain more realistic cask 

deceleration or equivalent load for quasi-static analyses. 

Alternatively, applicants can develop an analytical model to 

calculate cask deceleration loads. In the analytical approach, the 

hard receiving surface for a drop or tipover accident need not be 

an unyielding surface, and its flexibility may be included in the 

modeling. 

The structural discipline should review validation of the analytical 

model. The staff has completed a series of low-velocity impact 

tests of a steel billet from which a model validation approach and 

corresponding acceptance criteria have been developed. These 

tests and analytical evaluations are summarized in 

NUREG/CR-6608, Summary and Evaluation of Low-Velocity 

Impact Tests of Solid Steel Billet Onto Concrete Pads

(Witte, 1998). On the basis of the report, the following model 

validation acceptance criteria apply to a cask-pad-soil analytical 

model for predicting impact responses of the cask: 

 • When solid steel billet is used to replace the cask in the 

cask-pad-soil analytical model, it should predict a pulse 

amplitude slightly higher than the recorded pulse amplitude 

from the billet test. 

 • The calculated pulse duration and shape should be similar, 

but not necessarily identical, to those recorded from the 

billet test. 

The validated billet-pad-soil model is considered adaptable to a 

cask-pad-soil analysis model if relevant attributes of the cask are 

used to replace those of the billet. 

(b) Explosive Overpressure (LOW Priority) 
 3-17 

Explosion-induced overpressure and reflected pressure may result 

from explosion hazards associated with explosives and chemicals 

transported by rail or on public highways, natural gas pipelines, 

and vehicular fires of equipment used in the transfer of casks. 

Explosions may result from detonation of an air-gaseous fuel 

mixture. With the exception of transfer vehicle accidents, the 

explosion hazards are typically similar to those for facilities subject 

to reviews under 10 CFR Part 50, “Domestic Licensing of 

Production and Utilization Facilities.” 

The SAR should state the level of overpressure that the cask 

system can withstand for this accident condition. This 

overpressure level would then serve as the quantitative envelope 

for future comparison with hazards for specific site installations. 

The pressure criteria for the assumed design-basis wind or 

tornado may also serve as an envelope for the explosive 

pressures for comparison with actual site hazards of a general 

licensee’s facility. 

If the SAR includes bounding explosion effects for which the cask 

system is to be approved, the reviewer should verify that the 

applicant also provided structural analyses of those effects for 

cask system structures that may be affected. The SAR should 

identify the maximum response determined. The maximum 

response includes pressure-induced maximum stresses at critical 

cask locations and governing structural performance modes for 

the cask components important to safety. That response should 

be sufficiently low such that while damage may occur, it would not 

impair the capability of the component to perform its safety 

functions. In addition, the SAR should identify any post-event 

inspection and remedial actions that may be necessary. 

(c) Fire (LOW Priority) 

Chapter 4, “Thermal Evaluation” of this SRP addresses potential 

fire conditions. Fire-related structural evaluation considerations 

include increased pressures in the confinement cask, changes in 

material properties, stresses caused by different coefficients of 

thermal expansion and/or temperatures in interacting materials, 

and physical destruction. 

The reviewer should evaluate the discussion in the SAR 

concerning the treatment of structural effects associated with the 

presumed fire. The reviewer should evaluate the appropriateness 

of the applicant’s analysis of those structural effects for the 

assumed parameters of the design-basis fire. The reviewer 

should confirm that the applicant defined the confinement cask 

pressure capacity on the basis of the cask material properties at 

the temperature resulting from the fire. Spalling of concrete that 

may result from a fire is generally considered acceptable and 

need not be estimated or evaluated. Such damage is readily 
 3-18 

detectable, and appropriate recovery or corrective measures may 

be presumed. The NRC accepts concrete temperatures that 

exceed the temperature limits of ACI349 for accidents, providing 

that the temperatures result from a fire. However, corrective 

actions may need to be taken for continued safe storage. 

(d) Flood (LOW Priority) 

The applicant’s evaluation of the DSS design should be reviewed 

with regard to the structural consequences of a flood event. The 

SAR may stipulate an assumption that the DSS not be used at 

any site where there is potential for flooding. In this case, the 

DSS would have to be placed at an ISFSI site above the 

maximum probable flood level (SAR Chapter 12, “Accident 

Analyses Evaluation” should state this condition). Alternatively, an 

application for a certificate of compliance to use a DSS at a site 

with flooding potential would require a full analysis for a defined 

flood event. 

If a design flood event is defined for the certificate of compliance 

the reviewer should verify that the SSCs meet appropriate 

guidance in RG 1.59, Rev. 2 and 1.102, Rev. 1 for that level of 

flood protection. 

One possible structural consequence of a flood is that a vertically 

stored cask may tip over or translate horizontally (slide) because 

of the water velocity. Another possible consequence is that 

external hydrostatic pressure will exceed the capacity of the cask. 

The applicant may state the critical water velocity and hydrostatic 

pressure as bounds for the SAR flood analysis. 

The NRC accepts the evaluation for flooding events when the 

flood conditions for overturning and sliding of stored confinement 

casks and other cask system structures (with a safety factor of 1.1 

for accidents cases) have been applied. The applicant should 

state the basis for estimation of lateral pressure on a structure as 

a result of water velocity. 

The NRC accepts the use of Hoerner’s Fluid-Dynamics Drag

(Hoerner, 1965) for estimating drag coefficients and net lateral 

water pressure. An approach for calculating the velocity 

corresponding to the cask stability limit is to assume that the cask 

is pinned at the outer edge of the cask bottom and rotates about 

that outer edge, and the pinned edge does not permit sliding. The 

overturning moment from the velocity of the flood water can be 

compared to the stability moment of the cask (with buoyancy 

considered). The structural consequences of the flood event are 

typically bounded by analyses for the drop or tipover accident 

cases. 
 3-19 

The analysis of the confinement cask should be reviewed for 

flood-related hydrostatic pressure. The analysis should include 

the combined effects of weight, external hydrostatic pressure, 

internal pressure(s), and thermal stress. Resistance of the 

confinement cask to flood-related hydrostatic pressure should be 

analyzed in accordance with Section III, Subsection NB or NC, of 

the ASME B&PV Code (depending on the subsection used for 

design). 

Additional flood consequences include potential scouring under a 

foundation, damage to access routes, temporary blockage of 

ventilation passages with water, blockage of ventilation passages 

and interstitial spaces between the confinement cask and 

shielding structure with mud, and steep temperature gradients in 

the shielding structure and confinement cask. The consequences 

of these conditions may be analyzed in the SAR and identified in 

the certificate of compliance so a general licensee will be able to 

consider these factors when siting an ISFSI. 

(e) Tornado Winds (LOW Priority) 

The reviewer should verify that the SAR addresses the potential 

structural consequences of design-basis tornado or extreme wind 

effects. The load combination analyses should be reviewed for 

acceptable inclusion of tornadoes and tornado missiles. Current 

NRC guidance provided in RG 1.76, Rev. 1, recognizes three 

regions in the contiguous United States each with distinct designbasis tornado parameters. The applicant for a certificate of 

compliance must clearly define the boundary conditions of the 

proposed cask system with respect to these regions or utilized 

Region 1. 

Confinement casks may be vulnerable to overturning and/or 

translation caused by the direct force of the drag pressure while in 

storage or during transfer operations. Criteria for resistance to 

overturning or sliding should be provided in the SAR. 

Confinement casks are generally not vulnerable to damage from 

overpressure or negative pressure associated with tornadoes or 

extreme winds. However, they may be vulnerable to secondary 

effects, such as wind-borne missiles (see (f), below) or collapse of 

a weather enclosure, if used. The capability and behavior of the 

cask system under the collapse of any such external structure, if 

allowed by the Certificate of Compliance should be identified in 

the SAR. 

Tornadoes typically produce the greatest “design-level” wind 

effects for American sites. However, there are some potential 

American sites at which high winds may be more severe than the 

credible tornado. The SARs for a limited set of potential sites 

could reflect high wind effects as a basis for structural analysis. If 
 3-20 

the certificate is to include proven design resistance to tornadoes 

or extreme winds, the SAR documentation must identify the wind 

levels (e.g., in miles or kilometers per hour), source (tornado or 

high wind), and specific wind-driven missiles (shape, weight, and 

velocity) for which the design is to be evaluated. 

RG 1.76, Rev. 1, “Design-Basis Tornado for Nuclear Power 

Plants,” provides applicable tornado-related parameters. The 

NRC accepts the use of ASCE 7 for conversion of wind speed to 

pressure and for typical building shape factors. Conversion of 

tornado or other wind speeds to pressure in the SAR 

documentation should assume that the cask system is at sea 

level. 

The reviewer should verify that the cask system design meets 

appropriate guidance in the RG 1.76, Rev. 1, and 1.117, Rev. 1, 

and NUREG-0800 “Standard Review Plan for Power Reactors,” 

Section 3.3.2, Rev. 3 for tornado protection. 

Tornadoes and high winds can produce a significant negative 

pressure differential between interior spaces and the outside in a 

storage cask system that should be considered. This is a function 

of wind speed and factors relating to the structure. The magnitude 

of negative pressure depends on other parameters of the tornado 

or wind, and on wall pressure coefficients (as expressed in ASCE 

7). There is no need for the SAR to separately state negative 

pressure to establish an envelope for approval since negative 

pressure is insignificant with regard to confinement cask accident 

pressure analysis. 

The NRC does not accept the presumption that there will be 

sufficient warning of tornadoes that operations such as transfer 

between the fuel pool facility and storage site may never be 

exposed to tornado effects. Overturning during onsite transfer is 

considered by the staff to be a design-basis event. The tornado 

analysis should determine if tornado-induced overturning is 

bounded by drop and tipover cases. In addition, the SAR should 

show that the cask system will continue to perform its intended 

safety functions (i.e., criticality, radioactive material release, heat 

removal, radiation exposure, and retrievability). 

(f) Tornado Missiles (LOW Priority) 

The applicant’s evaluation of the cask system design should be 

reviewed with regard to the structural consequences of winddriven missile impact (RG 1.76, Rev. 1 and NUREG-0800, 

“Standard Review Plan for Power Reactors,” Section 3.5.1.4 

(Rev. 3) and Section 3.5.3 (Rev. 3) describe the effects of tornado 

missiles). The SAR should define the missile parameters for 

which the cask system is to be evaluated based on the three 

tornado regions currently identified in the RG 1.76, Rev. 1. 
 3-21 

Among the possible missile effects, the SAR should address those 

that may result in a tipover and those that may cause physical 

damage as a result of impact. The damage should not result in 

unacceptable radiation dose or significantly impair either criticality 

control, heat removal, or the retrievability of the fuel. 

The NRC has accepted use of the analytical approaches given in 

U.S. Reactor Containment Technology, ORNL-NSIC-5, Volume 1, 

Chapter 6 (Cottrell and Savolainen), for estimating the potential 

effects of missile impact on steel sheets, plates, and other 

structures. Further guidance on analytical acceptable approaches 

for use in ISFSI design is provided in NUREG-0800, Section 3.5.3, 

“Barrier Design Procedures.” In addition, for analysis and design 

regarding the ability of reinforced concrete structures to resist 

missiles, the NRC has accepted use of “Review of Procedures for 

the Analysis and Design of Concrete Structures to Resist Missile 

Impact Effects” (Kennedy, 1975). 

Cask systems are not required to survive missile impacts without 

permanent deformation. However, the maximum extent of 

damage from a design-basis event must be predicted and should 

be sufficiently limited. Moreover, the capability of the SSC to 

perform their safety functions should not be impaired. 

(g) Earthquake (MEDIUM Priority) 

The applicant’s evaluation of the cask design should be reviewed 

with regard to the structural consequences of the earthquake 

event. Cask designs must satisfy the load combinations that 

encompass earthquake, including those for sliding and 

overturning. The applicant should demonstrate that no tipover or 

drop will result from an earthquake. In addition, impacts between 

casks should either be precluded, or should be considered an 

accident event for which the cask must be shown to be structurally 

adequate. 

Appendix H of ANSI/ANS-57.9-1992 provides guidance for 

seismic analysis. Implicit in this guidance is the assumption that 

the ISFSI concrete pad, supported by soil, behaves as a rigid mat 

and therefore possesses no out-of-plane flexibility. This is valid 

for the majority of nuclear power plant structures where relatively 

thick mats support integral reinforced concrete walls. However, 

ISFSI pads are usually relatively thin structures (i.e., small 

thickness to length ratio) and generally do not incorporate integral 

walls to stiffen the pad. While the cask itself is relatively rigid, the 

rigid cask resting on a flexible pad has a lateral mode frequency 

that is generally low enough to fall within the amplified range of 

most design earthquake spectra. Thus, in determining the inertia 

forces that act at the center of gravity of the cask for the purpose 

of evaluating the onset of sliding or tipping, the reviewer should 

ensure that the applicant has either accounted for the out-of-plane 
 3-22 

flexibility of the pad in the seismic analysis or demonstrated that it 

is not an important parameter in determining the response of the 

cask, (“Influence of ISFSI Design Parameters on the Seismic 

Response of Dry Storage Casks,” Bjorkman & Moore, 2001). 

The reviewer should verify that the cask system design meets 

appropriate guidance in RGs 1.29, Rev. 4, 1.60, Rev. 1, 1.61, 

Rev. 1, and 1.92, Rev. 2, for protection against seismic events. 

The SAR documentation should include analysis of the potential 

for impacts between components of the cask system. These 

could include contact between the confinement shell and its inner 

components or outer shield and the rocking and fall back of a 

vertically or horizontally oriented confinement cask on its supports. 

Cask systems are not required to survive a design earthquake 

without permanent deformation. However, the maximum extent of 

damage from a design earthquake must be predicted, and the 

capability to provide principal safety functions should not degrade. 

ii. Structural Analysis Methods 

(LOW Priority) The applicant’s structural analysis of various loading combinations 

and the resulting stresses, strains, and deformations from different loads should 

be reviewed. The reviewer should verify that the applicant properly used 

acceptable analytical approaches and tools. In addition, the applicant should 

have performed and reviewed the associated computations internally under an 

acceptable independent design review (equivalent to ASME NQA-1) and quality 

assurance procedures. The scope of the staff’s review may include performing 

detailed parallel computations (such as finite element analyses) to validate 

submitted computations or their results. The reviewer may perform separate, 

less extensive calculations when these could most readily evaluate any 

suspected problems. 

The applicant’s analysis of loads and load combinations resulting from different 

structural conditions should be consistent with the code or criteria requirements 

used in designing the component. 

Subsection NB or NC of the ASME B&PV Code defines the requirements for 

categorizing stresses and determining allowable stress limits for the confinement 

boundary of the cask. For the fuel basket, Subsection NG of the Code applies. 

These references also provide definitions of stress categories and stress 

intensity limits for normal and off-normal operating conditions. For Level D or 

accident conditions, Appendix F to the ASME B&PV Code provides definitions of 

the stress intensity limits. 

In accordance with these references, stress intensity is defined on the basis of 

the maximum shear stress theory for ductile materials. Since the maximum 

shear stress is not identical to the maximum octahedral shear stress, octahedral 

shear stresses should not be compared with the stress intensity limits. Values 

for the stress intensity limits are defined in Appendices I and III of the ASME 
 3-23 

Code. Stresses resulting from inertial and pressure loads should be considered 

primary stresses. Thermal stresses resulting from temperature gradients may be 

considered secondary stresses if they are self-limiting and do not cause 

structural failure. Stresses due to thermal gradients in fuel baskets may not be 

self-limiting and should be considered by the applicant because of the possibility 

of uneven heat loadings of adjacent assemblies as well as the effects of 

asymmetry in the basket structure. 

(1) Finite-Element Analyses (HIGH Priority) 

Because of the complexity of many structural design considerations and 

load conditions, structural design computations are often performed using 

finite-element analysis. 

The applicant should perform the finite-element analyses using a generalpurpose program that is well benchmarked and widely used for many 

types of structural analyses. 

Consistent with the provisions of ASME Code, Section III, Appendix F, 

inelastic material properties may be used for the storage cask design 

analysis evaluation for accident loads. The SAR should identify the 

sources used for the inelastic material properties. 

Lead shielding can be modeled either with elastic or inelastic properties. 

The elastic modulus and limit used for lead in the elastic analysis should 

be determined on the basis of the potential temperature of the material. 

An appropriate plasticity model of lead can be used to account for its 

inelastic behavior. 

Nonstructural components of the confinement cask are generally not 

included in finite element models. However, the models should include 

any influence these nonstructural components may have on the structural 

performance of the cask. Possible influences include the nonstructural 

components’ inertial weight, restraint to motion of the structural 

components, and localized influence on load applications because of 

geometrical effects. 

Bolted connections can be modeled either discretely or with contact 

conditions. To discretely model the bolted connections, the applicant 

should use appropriate element types and material properties. With 

contact conditions, the interfaces joined by the bolts can be modeled as 

tied. 

Verify that the applicant has provided information on any computer-based 

modeling as described in Appendix 3A to this chapter, and review the 

structural analyses submitted by the applicant in accordance with the 

Appendix. 

(2) Closed-Form Calculations (MEDIUM Priority) 
 3-24 

The applicant should perform closed-form calculations for relatively 

simple structural load conditions or conditions for which a formula has 

been developed. Closed-form calculations are also typically used to 

check the results of finite-element analyses. In addition, this type of 

calculation can be used for analyses involving principles of conservation 

of energy and comparisons of overturning moments. 

One source of closed-form equations accepted by the NRC is Formulas 

for Stress and Strain (Roark, 1965). Use of a particular equation or 

formulation for the load conditions should be justified. The most 

important aspect of the calculations to evaluate is the basis for the 

assumptions used in the calculations. In many cases, the calculations 

are faulty in that they fail to include portions of the cask, or the load 

conditions are idealized inappropriately. 

To be consistent with the provisions in Section III of the ASME Code, the 

analyses should use linear material properties. Linear analysis should be 

the basis for all closed-form calculations. 

(3) Structural Analysis for Specific Cask Components 

The following paragraphs present a few specific examples of structural 

analysis for some of the confinement cask components of a cask storage 

system. 

(a) Fuel Basket (HIGH Priority) 

The fuel basket design should be reviewed to assess the 

applicant’s analysis of the combined effects of weight, thermal 

stresses, and cask-drop impact forces that could arise during 

spent fuel transfer and storage operations. The weight supported 

by the basket should be the maximum or design weight of the 

SNF to be stored. In addition, the applicant should evaluate all 

credible potential orientations of the cask and basket during cask 

transfer and handling drops while transferring the spent fuel into 

storage. End or side drops typically produce the greatest 

structural demand on various basket components. In an end drop, 

the basket is supported by the bottom of the confinement cask 

cavity upon impact. In the side drop, the basket structure and 

points of contact with the confinement cask must support the 

mass of the basket and loaded fuel. 

In previous DSS evaluations, the NRC has accepted two 

approaches for analyses regarding the structural capability of the 

basket to acceptably survive a cask drop during transfer and 

storage. The first approach uses dynamic analyses in a two-step 

process. In Step 1, the applicant performs a dynamic analysis of 

the cask body impacting a target surface and assesses the 

performance of the cask body, including determining the timehistory response from the cask drop impact. In Step 2, this timehistory response can be translated into a forcing function that can 
 3-25 

be applied to the supporting contact points of an appropriate 

model of the fuel basket. 

The second approach uses a quasi-static analysis of the basket 

subjected to the equivalent acceleration inertial load derived from 

the cask-drop impact analysis. In this analysis, the applicant 

should apply the equivalent acceleration inertial load using an 

appropriate model of the basket with the location(s) most 

vulnerable to the impact. Support provided by the inside surface 

of the cask cavity should be represented by the appropriate 

boundary conditions on the outside edge of the basket. In 

addition, the applicant should conservatively select the equivalent 

acceleration inertial load such that it bounds the possible inertial 

loads resulting from a cask-drop accident onto the bounding target 

surfaces. If applicable, the inertial load should also account for 

dynamic amplification effects by using a dynamic amplification 

factor. 

The applicant should also evaluate the buckling capacity of the 

cask basket materials. Acceptable guidance for this evaluation is 

provided in Section III of the ASME B&PV Code and NUREG/CR6322, “Buckling Analysis of Spent Fuel Basket,” (Lee and 

Bumpas, 1995). For this evaluation, the applicant should select 

the appropriate end conditions used in the buckling capacity 

equations on the basis of sensitivity studies. These studies can 

bound the range of conditions that are typically either fixed for a 

welded connection or free if there is no rigid connection. 

(b) Closure Lid Bolts of Confinement Boundary (MEDIUM Priority) 

The design analysis for the closure-lid bolts should be reviewed to 

ensure that it properly includes the combined effects of weight, 

internal pressure(s), thermal stress, O-ring compression force, 

cask impact forces, and bolt pre-load. Typically, applicants 

specify the pre-load and bolt torque for the closure bolts on the 

basis of bolt diameter, and the coefficient of friction between the 

bolt and the lid. Externally applied loads (such as the internal 

pressure and impact force) produce direct tensile force on the 

bolts as well as an additional prying force caused by lid rotation at 

the bolted joint. The tensile bolt force obtained by adding together 

the pressure loads, impact forces, thermal load, and O-ring 

compression force should then be compared with the tensile bolt 

force computed from the pre-load and operating temperature load 

alone. The larger of the two calculated tensile forces should 

control the design. The maximum design bolt force should then 

be obtained by combining the larger direct tensile bolt force with 

the additional prying force. The weight is derived from the 

maximum or design weight of the closure lids and any cask 

components supported by the lids. Acceptable analytical methods 

for closure bolts are given in NUREG/CR-6007, “Stress Analysis 

of Closure Bolts for Shipping Casks” (Mok and Fischer, 1993). 
 3-26 

The bolt engagement lengths should be reviewed. If the lids are 

fabricated from relatively non-hardened materials, threaded 

inserts may be used in the closure lids to accommodate the 

hardened material of the bolts. 

 (c) Trunnions (LOW Priority) 

The design of the trunnions, their connections to the cask body, 

and the cask body in the local area around the trunnions should 

be reviewed. The design basis for the trunnions can be either 

non-redundant or redundant. In either case, the design should 

meet the requirements of ANSI N14.6 for critical loads and the 

requirements of NUREG-0612, “Control of Heavy Loads at Power 

Plants.” 

Non-redundant lifting systems should be designed for not less 

than 6 times the material yield strength and 10 times the material 

ultimate strength given the design lift weight of the loaded cask. 

Redundant lifting systems should be designed for not less than 

3 times the material yield strength and 5 times the material 

ultimate strength given the design loaded lift weight of the cask. 

Acceptance testing requirements for trunnions are discussed in 

Chapter 10, “Acceptance Tests and Maintenance Program 

Evaluation,” of this SRP. 

For a typical trunnion design, the maximum stress occurs at the 

base of the trunnion as a combination of bending and shear 

stresses. A conservative technique for computing the bending 

stress is to assume that the lifting force is applied at the 

cantilevered end of the trunnion, and that the stress is fully 

developed at the base of the trunnion. If other assumptions, 

including ASME Section III stress limits by the finite element 

design analysis and slight material yielding at localized regions, 

are considered, the applicant should provide adequate 

justifications. 

iii. Structural Evaluation 

(1) Structural Capability (LOW Priority) 

The applicant’s structural analyses should be reviewed to assess the 

information regarding margins of safety or compliance with ASME Code 

stress limits, overturning margins, and other criteria appropriate for the 

division of the ASME Code being used. The comparisons of capability 

versus demand for the various applicable loading conditions should be 

presented in the same terms used in the design code (e.g., type of 

stress). In addition, margins of safety should be included on the basis of 

comparisons between capacity and demand for each of structural 

component analyzed. The minimum margin of safety for any structural 
 3-27 

section of a component should be included for the different load 

conditions. 

(2) Fabrication and Construction (MEDIUM Priority) 

The NRC has accepted fabrication of metallic confinement casks in 

accordance with Section III, Division 1 of the ASME B&PV Code. If the 

fabrication, construction, or assembly deviate in any way from the 

subsection of this standard used for design, the SAR must explicitly state 

the applicant’s justification for the deviation, and the justification must be 

acceptable to the NRC. 

If the design of the confinement cask is proposed to be governed by 

ASME, Section III, Division 2, similar to a metallic-lined concrete pressure 

vessel NRC would expect the fabrication/construction of such a cask to 

also be governed by the Division 2 requirements. Any deviations from the 

Code requirements should be addressed as noted for Division I above for 

metallic containment. 

If the design of the confinement cask is proposed to be governed by 

ASME, Section III, Division 3, the applicant will have to provide 

supplemental details to the Code provisions since Subsection WC does 

not provide guidance to address all construction details for classic 

containments. 

3.5.2 Other System Components and Structures Important to Safety 

3.5.2.1 Scope 

This portion of the DSS structural review provides guidance by addressing procedures for 

evaluating all structures that are important to safety (as defined in 10 CFR Part 72.3), whether 

steel, concrete or other material not addressed as the confinement cask and internals 

(Subsection 3.5.1). Structures may include items such as gamma and neutron shielding, 

overpack material, any respective encasement foundations, structural supports, ventilation 

passages, weather enclosures, earth retention structures, and protective structures. This 

evaluation should include drawings, plans, sections, and technical specifications for these 

SSCs. 

3.5.2.2 Structural Design Criteria and Design Features 

 i. Design Criteria (MEDIUM Priority) 

 (1) General Structural Requirements 

Structural requirements are driven by the functional roles of the system 

components and the need to maintain safety. Safety requirements are 

expressed in the referenced rules, standards, and codes and as criteria 

specific to the component. The basic safety requirements are that the 

structural and functional design must preclude the following: 

 • Unacceptable risk of criticality. 
 3-28 

 • Unacceptable release of radioactive materials to the environment. 

 • Unacceptable radiation dose to the public or workers. 

 • Significant impairment of retrievability of stored nuclear materials 

during normal and off-normal conditions. 

 The applicant should consider the potential for liquefaction and other soil 

instabilities attributable to vibrating ground motion, for any structure or 

system component such as a cask system support pad. 

 Reinforced concrete pads that support confinement casks in storage do 

not constitute “pavements.” As such, they should be designed and 

constructed as foundations under an applicable code such as, ACI 349, 

ACI 318, or IBC. Such pads typically are not classified as important to 

safety; however, in some cases they may be. 

 Steel embedments in reinforced concrete structures must satisfy the 

requirements of the design code applicable to the reinforced concrete 

structure. Similarly, structural steel must satisfy the requirements of the 

applicable steel design code (e.g., ASME B&PV Code, AISC, or other 

identified code). 

 (2) Applicable Codes and Standards 

The codes and standards identified in the SAR should be reviewed as 

well as their proposed applications. This subsection addresses the codes 

and standards that the NRC has accepted for structures important to 

safety categorized by application that are not confinement casks or the 

steel internals. 

The NRC accepts the use of ANSI/ANS-57.9 (together with the codes and 

standards cited therein) as the basic reference for the structures 

important to safety that are not designed in accordance with the Section 

III, Division 1 or Division 2 of the ASME B&PV Code. However, both the 

lifting equipment design and the devices for lifting system components 

that are important to safety must comply with ANSI Standard N14.6. The 

NRC accepts the load combinations shown in Table 3-3 for structures not 

designed under either Section III of the ASME B&PV Code Section III, 

Division 1 or 2 (ACI 359). See Table 3-2 for loads and their descriptions. 

The reviewer should review the suitability of the applicant’s identification 

of codes and standards that are to be met by the structural design and 

construction of other components subject to NRC approval. The principal 

codes and standards include the following references that may apply to 

steel structures and components as well as concrete portions of the cask 

system: 

 • AISC, “Specification for Structural Steel Buildings – Allowable 

Stress Design and Plastic Design.” The NRC has not yet received 
 3-29 

any applications that propose a steel design on the basis of the 

AISC’s “Load and Resistance Factor Design (LRFD) Specification 

for Structural Steel Buildings.” If such a design was received, the 

NRC would evaluate the proposal for compliance with the load 

combinations summarized in Table 3-3 and for consistent 

application of the LRFD design methodology. 

 • To date, the NRC has not required applicants to design or build 

structural steel components of a cask system important to safety 

in compliance with ANSI/ANS N690, “Nuclear Facilities — Steel 

Safety-Related Structures for Design Fabrication and Erection.” 

 • AWS D1.1, “Structural Welding Code Steel.” 

C ASCE 7, “Minimum Design Loads for Buildings and Other 

Structures.” 

 • ACI 349, Appendix D, for anchoring to concrete or Section 10.14 

for composite compression sections, as applicable, when 

constructed of structural steel embedded in reinforced concrete. 

Where requirements do not conflict, the steel must also comply 

with the requirements of the codes stated above. In addition, ACI 

349 defines constraints for obtaining ductile response to extreme 

loads by ensuring that the strength of steel embedments controls 

the design; these constraints must not be subverted by overdesign of the steel. 

 • For reinforced concrete the NRC has not accepted the use of a 

set of criteria selected from multiple standards and codes, except 

when the selected criteria meet the most limiting requirements of 

each code. However, in recognizing a graded approach to quality 

assurance, the NRC has approved the use of ACI 349 for design 

and material selection for reinforced concrete structures important 

to safety (not confinement). The NRC has allowed the optional 

use of ACI 318 as an alternative standard for construction as 

described below. 

 • In both cases, however, the design, material selection and 

specification, and construction must also meet any additional or 

more stringent requirements given in ANSI/ANS-57.9. 

The following paragraphs identify the portions of ACI 349 that 

apply to design (including material selection) and must be met by 

applicants who choose to use ACI 318 for construction. (The 

paragraph references are as in ACI 349-06.). Unlisted and 

excepted sections address construction requirements for which 

the NRC accepts substitution of ACI 318. 

Chapter 1 “General Requirements,” Sections 1.1 and 1.5 

(except references to construction), and Sections 

1.2 and 1.4. 
 3-30 

 Chapter 2 “Definitions.” 

 Chapter 3 “Materials” (except Sections 3.1, 3.2.3, 3.3.3, 

3.5.3.1.1, 3.6.1.0, and 3.7). 

 Chapter 4 ”Durability Requirements” 

 Chapter 6 “Form Work, Embedded Pipes, and Construction 

Joints,” Sections 6.3.13, 6.3.14, and 6.3.15. 

 Chapter 7 “Details of Reinforcement.” 

 Chapter 8 “Analysis and Design General Considerations.” 

 Chapter 9 “Strength and Serviceability Requirements.” 

 Chapter 10 “Flexure and Axial Load.” 

 Chapter 11 “Shear and Torsion.” 

 Chapter 12 “Development and Splices of Reinforcement.” 

 Chapter 13 “Two-way Slab Systems.” 

 Chapter 14 “Walls.” 

 Chapter 15 “Footings.” 

 Chapter 16 “Precast Concrete.” 

 Chapter 17 “Composite Concrete Flexural Members.” 

 Chapter 18 “Prestressed Concrete.” 

 Chapter 19 “Shells.” 

 Appendix A “Strut-and-Tie Models.” 

 Appendix D “Anchoring to Concrete.” 

Appendix E “Thermal Considerations.” 

 Appendix F “Special Provisions for Impulsive and Impactive 

Effects” (except that the load combinations included 

herein, must be used. 

For fluid systems used with a cask system that may be connected 

to a penetration of the confinement barrier outside an enclosing 

structure licensed under 10 CFR Part 50 (e.g., the fuel pool 

building), the NRC accepts construction consistent with 

requirements comparable to those used for Quality Group C, as 

shown in RG 1.26, “Quality Group Classifications and Standards 

for Water-, Steam-, and Radioactive Waste-Containing 

Components of Nuclear Power Plants,” Rev. 4 and 

NUREG-0800,” Section 3.2.2, “Standard Review Plan for Nuclear 

Power Plants.” In this context, “construction” includes materials, 

design, fabrication, examination, testing, inspection, and 

certification required in the manufacture and installation of 

components. Quality Group D may, under some circumstances 

be justified. 

Quality Group C requires construction of piping, pumps, valves, 

atmospheric storage tanks, and 0-15 psig storage tanks in 

conformance with Section III of ASME B&PV Code 1, Class 3 

(Subsection ND). In addition, Quality Group C requires that 

supports for these components meet the requirements of 

Subsection NF. 

By contrast, Quality Group D requires compliance with the 

following codes, as a minimum: 
 3-31 

Piping: ANSI/ASME B31.1, “Power Piping.” 

Pumps: Manufacturer’s Standards. 

Valves: ANSI/ASME B31.1 and ANSI B16.34, “Valves.” 

Atmospheric Storage Tanks: 

American Water Works Association (AWWA), 

“Standard for Steel Tanks — Standpipes, 

Reservoirs, and Elevated Tanks for Water Storage” 

(AWWA D100) or ANSI/ASME B96.1, “Specification 

for Welded Aluminum-Alloy Field-Erected Storage 

Tanks.” 

0–15 psig Storage Tanks: 

American Petroleum Institute’s (API) 

“Recommended Rules for Design and Construction 

of Large, Welded, Low-Pressure Storage Tanks” 

(API 620). 

The NRC accepts the “Boundaries of Jurisdiction” applicable to 

Section III, Subsections NB-1130 and NC-1130, of ASME B&PV 

Code. These boundaries apply to attachments to penetrations of 

the confinement barrier outside an enclosure licensed under 10 

CFR Part 50. Specifically, these boundaries define whether the 

attachments must be designed, fabricated, and installed in 

accordance with Section III, Subsection NB or NC, of ASME 

B&PV Code. 

Note that codes, other than those discussed herein (e.g., the 

“Electric, Life Safety, and Lightning Protection Codes” 

promulgated by the National Fire Protection Association [NFPA]), 

may apply to the design and construction of the cask system. It is 

acceptable to include such codes in the design by inclusion in the 

SAR. Where designs of structures subject to approval are also 

covered by such other codes, the review should include evaluation 

of compliance with those codes. 

The NRC has not yet received any applications for licensing or 

approval of a cask system that included masonry important to 

safety. Masonry is not considered suitable for confinement, but it 

may be acceptable for enclosures and physical or radiationshielding applications. 

ii. Structural Design Features (MEDIUM Priority) 

The design description in the SAR documentation should be reviewed to ensure 

that it defines the functional performance required of the structures. The design 

description of the non-confinement safety-related structures of the cask system 

should provide a clear understanding to be reached by the reviewer of the 

significance of the safety-related features to the required performance. 
 3-32 

The SAR documentation should also be reviewed regarding the physical design 

of the structures important to safety. This should include the following as a 

minimum. As appropriate to the specific structure the following information 

should be provided. 

• Dimensioning of all structural elements. 

• Locations, sizes, configuration, spacing, welding, fasteners etc. of the 

safety-related non-confinement structures should be provided. 

• Locations and specifications for controls, that will be necessary in 

fabrication and construction. 

• Structural materials with defining standards or specifications summarized 

or references to Chapter 8, “Materials Evaluation” of this SRP herein 

should be reviewed. 

• Information on the physical design of attachments, embedments, and 

other structural elements should be provided. 

Auxiliary cask system equipment important to safety has often been specially 

designed. In particular, the structural design features that provide for safety 

should be supported by design or operational analysis. This analysis should 

demonstrate that the equipment will meet the basic safety criteria, regardless of 

problems that may occur in mechanical, electrical, human operator, or other 

operations. 

The NRC has accepted and approved cask system designs that depend on the 

operation of new mechanical systems for system use. NRC approval does not 

certify that the mechanical systems will operate as projected but rather that 

proper functioning is necessary to successfully complete a specified operation. 

Such approval reflects a finding by the NRC staff that, regardless of the system’s 

success (or lack thereof) in mechanical operation, the basic safety criteria will be 

met, as stated above. 

The proposed system design should be reviewed against planned normal and 

off-normal, operations and accidents. The reviewer should determine whether 

the structural design of the equipment provides for continuing satisfaction of the 

basic safety criteria. The reviewer should consider that the equipment could fail 

to operate at any time (i.e., during operations at the physical limits of speed or 

range, or during a credible, off-normal, or accident-level event). 

3.5.2.3 Structural Analysis 

Subsections 3.5.1.4 (i) and (ii) provide guidance regarding structural analysis for the 

confinement cask and metallic internals of cask systems. These subsections provide 

supplemental guidance primarily related to steel and concrete structures, other than the 

confinement cask and its contents and integral components that are important to safety. The 

appropriateness, completeness, and correctness of the applicant’s proposed implementation of 
 3-33 

these load conditions and combinations for the metallic and reinforced concrete structures 

should be reviewed. 

i. Load Conditions (MEDIUM Priority) 

The load definitions and combinations shown in Tables 3-2 and 3-3 have been 

accepted by the NRC for analysis of steel and reinforced concrete ISFSI 

structures that are important to safety. These load combinations are included in 

or derived from ANSI/ANS 57.9 and ACI 349. 

Structures that are important to safety should have sufficient capability for every 

section to withstand the worst-case loads under normal and off-normal 

conditions. Such capability ensures that these structures will not experience 

permanent deformation or degradation of the capability to withstand any future 

loadings. 

The NRC accepts the load combinations in Table 3-3 that implement and 

supplement those of ANSI/ANS-57.9. 

(1) Normal Conditions 

The SAR documentation should be reviewed to ensure adequate 

inclusion of the following conditions that may be of particular concern for 

concrete structures important to safety if the loading condition is 

appropriate: 

• Live and dynamic loads associated with transfer of the 

confinement cask to and from its storage position and in its 

storage location for its service lifetime. 

• Live and dynamic loads associated with installing closures. 

• Load or support conditions associated with potential differential 

settlement of foundations over the life of the cask system. 

• Thermal gradients associated with the normal range of operations 

and ranges of ambient temperature. 

• Thermal gradients that may result from impingement of 

precipitation on highly heated concrete. 

(2) Off-Normal Conditions 

The SAR should be reviewed to ensure adequate inclusion of the 

following off-normal operations and events: 

• Live and dynamic loads associated with equipment or instrument 

malfunctions, or accidental misuse during transfer of the 

confinement cask to and from its storage position. 
 3-34 

• Situations in which a confinement cask is jammed or moved at an 

excessive speed into contact with a reinforced concrete structure. 

• The impact of reinforced concrete structures by a suspended 

transfer, confinement, or storage cask. 

• Off-normal ambient temperature conditions (although they may be 

less severe than accident conditions, these may be of concern 

because of different sets of factors in the off-normal and accident 

load combinations, and because concrete temperature limits for 

off-normal conditions are the same as for normal conditions. Note 

that greatly elevated concrete temperatures are allowed for 

accident conditions in accordance with ACI 349, Section A.4). 

(3) Accident Conditions and Natural Phenomena Events 

The SAR should be reviewed for adequate inclusion of the following 

conditions associated with accident and conditions that may be of special 

concern for reinforced concrete structures: 

• Loads associated with accidental drops or other impacts during 

transfer of the confinement cask to and from its storage position. 

• Events that produce extreme thermal gradients in the concrete. 

• Contact caused by earthquake between the confinement cask and 

the reinforced concrete structures. 

• Drop of a closure into position or onto the structure. 

The ACI codes are intended to ensure ductile response beyond initial 

yield of structural components. ACI 349 also imposes conditions on 

design (beyond those of ACI 318) that effectively increase ductility. In 

particular, the reviewer should review the proposed reinforced concrete 

design to ensure that it provides code levels of ductility by satisfying the 

pertinent ACI 349 provisions. Seismic loads are considered to be 

“impulsive” and, therefore, are subject to the additional design constraints 

of Appendix F to ACI 349. Other accident conditions or natural 

phenomenon events may also produce impulsive or impactive loadings 

requiring the additional requirements of Appendix F to ACI 349. 

Reviewers should check the steel reinforcement schedules and drawings 

to ensure that any reinforcing steel quantities, sizes, and locations are 

consistent with the design analysis. 

In particular, consider the following aspects of the design: 

• Upper limit (60 ksi, 4219 kgf/cm2) on the specified yield strength 

of reinforcement, lower limit (3 ksi, 211 kgf/cm2) on concrete 

specified compressive strength (f”c), and upper limit on concrete 
 3-35 

strength, as analyzed and specified for the ISFSI cask storage 

pads. 

• Limit on the amount (cross-section area) of compressive 

reinforcement in flexural members. 

• Requirements on continuation and development lengths of tensile 

reinforcement. 

• Specifications for confinement and lateral reinforcement in 

compression members, in other compressive steel, and at 

connections of framing members. 

• Aspects of the design that ensure flexure controls (and limits) the 

response. 

• Requirements for shear reinforcement. 

• Limitations on the amount of tensile steel in the flexural members 

relative to that which would produce a balanced strain condition. 

• Projected maximum responses to design-basis loads within the 

permissible ductility ratios for the controlling structural action. 

• Embedments designed for ductile failure and to fail in the steel 

before pullout from the concrete. 

In addition, the construction specifications or descriptions (to the 

extent included in the SAR documentation) should be reviewed to 

ensure that substitution of materials, use of larger sizes, or 

placement of larger quantities of steel will be precluded, and that 

provisions for splicing or development of reinforcing steel will not 

reduce ductility of the members. 

ii. Structural Analysis Methods (HIGH Priority) 

The applicant should select and use analytical methods that are appropriate for 

the proposed type of materials and construction. In certain instances, however, 

the applicant may have to adapt existing analytical methods, codes, and models 

for highly specialized cask system equipment designs. Such instances require 

special review attention. In particular, the reviewer should ensure that the 

adapted approach is fully documented, supported, and acceptable. In addition, 

the reviewer should consider the potential for safety-related risk associated with 

a possible error in the design of special cask system equipment. The degree of 

risk indicates the suitability and acceptability of the adapted approach. 

Subsection 3.5.1.4.ii provides acceptable analytical methods of analysis that can 

be utilized. Appendix 3A addresses the application of computational modeling 

software. 

iii. Structural Evaluation (LOW Priority) 
 3-36 

In evaluating the variety of cask system equipment and structures that may be 

important to safety, the reviewer should ensure compliance with the basic safety 

criteria in Subsection 3.5.2.2 (i)(1) and that the specified parameters for 

acceptability such as stress, strain or deflection are within the permitted values 

identified in Subsection 3.5.2.2.i.(2). 

The NRC accepts strength design as presented in the current revision of ACI 349 

for reinforced concrete structures important to safety that are not within the scope 

of ACI 359. If the applicant uses another design approach, the review conducted 

within the scope of the DSS SAR evaluation should include in-depth comparison 

of that approach with the provisions of ACI 349. 

The NRC accepts the use of guidance in NUREG-0800 for analysis of natural 

phenomena, as related to the conditions that apply to the design of cask 

systems. However, the load combinations shown in Table 3-3 and the design 

and construction requirements of the codes cited above take precedence. The 

NRC accepts the American Society of Civil Engineers’ “Seismic Analysis of 

Safety Related Nuclear Structures” (ASCE 4) and ASCE 7 as the standards for 

seismic analysis. In addition, the NRC accepts tornado missile impact analysis in 

accordance with Kennedy’s Review of Procedures for the Analysis and Design of 

Concrete Structures to Resist Missile Impact Effects.

(1) Structural Capability (LOW Priority) 

 

Section 3.5.1.4.iii (1) addresses the assessment of the structures 

capability with respect to the ASME Code stress limits which are 

appropriate for metallic structures under Division 1 and for concrete 

structures under Division 2. 

 

For other safety related structural concrete, strength (or “ultimate 

strength”) design is the approach usually used in reinforced concrete 

design. Strength design is the only design approach that has been 

accepted for reinforced concrete structures that are part of cask systems 

not within the scope of ACI 359, and it is the approach used in the current 

revisions of ACI 349. This design code was tested and developed on the 

basis of extensive empirical experience with concrete construction. The 

current strength design approach, as presented in this code, includes 

empirically derived requirements and constraints. Determination that a 

reinforced concrete structure designed by another approach satisfies 

ACI 349 typically requires clause-by-clause review of the code for 

compliance. Allowable stress design was formerly used as the basis for 

ACI codes related to reinforced concrete design. However, those codes 

do not reflect additional experience gained through observations of 

structural performance and experimental testing that has since been 

included in the current approach to strength design. 

 

With respect to structural steel or other metallic structures important to 

safety, but not to the confinement structure or internals, the structural 

capability of the design may be based on the ASME Code with the use of 

the appropriate subsections as identified in Section 3.5.2.2 (i)(2) herein, 

or the AISC specifications also identified. Allowable stress, plastic 
 3-37 

design, and load and resistance factor methods of design are acceptable 

for use when there is justification for the method used provided in the 

application. 

 (2) Fabrication and Construction (MEDIUM Priority) 

For structures and structural components analyzed and designed based 

on ASME B&PV Code requirements of Section III, Division 1 or 

Division 2, the fabrication and construction provisions of these documents 

should form the basis for the production and installation of the structures 

and components of the cask storage system. 

NRC accepts construction in accordance with ACI 349 or ACI 318. 

Selection and validation of the proper concrete mix to meet design 

requirements are considered a construction function. By contrast, 

specification of cement type, aggregates, and special requirements for 

durability and elevated temperatures is considered a design or material 

selection function and is, therefore, governed by ACI 349 (and/or ACI 

359, if applicable). 

The following sections of ACI 318 (chapters, appendix, and 

paragraphing per ACI-318-02) have been accepted by the NRC 

for construction of ISFSI reinforced concrete structures that are 

not within the scope of ACI 359: 

Chapter 1 “General Requirements,” Sections 1.1.1, 1.1.2, 

1.1.3, and 1.1.5 (except references to design and 

material properties), and Section 1.3. 

Chapter 2 “Definitions” (use ACI 349, Chapter 2). 

Chapter 3 “Materials,” Sections 3.1 and 3.8 (except A-616, 

A-617, A-767, A-775, A-884, and A-934). 

Chapter 4 “Durability Requirements.” 

Chapter 5 “Concrete Quality, Mixing, and Placing.” 

Chapter 6 “Form Work, Embedded Pipes, and Construction 

Joints” (except references to design and material 

properties, which are governed by ACI 349). 

3.5.3 Other Structural Components Subject to NRC Approval (MEDIUM Priority) 

3.5.3.1 Scope 

The cask system description provided in the SAR may include a variety of components that are 

not important to safety such as transporters, ram systems, vacuum drying systems, drain and fill 

quick disconnects, support pads and other concrete structures not important to safety. These 

components should be reviewed to ensure proper functioning to the extent that the structures 

represent required elements of the total cask system. In particular, the reviewer should 

evaluate all structures that are proposed for approval in a cask system design acceptable to the 

NRC. This evaluation should ensure that the SAR provides sufficient information to confirm the 

proper functioning of the components and the overall system. For each system element that is 

not important to safety, the reviewer should address the potential response to accidents and 
 3-38 

natural phenomenon events to ensure that the given element will not jeopardize the safety 

provided by other system elements. 

3.5.3.2 Structural Design Criteria and Design Features 

i. Design Criteria 

(1) General Structural Requirements 

Structures subject to approval but not important to safety should be 

reviewed on the basis of determining whether the structures can properly 

perform their intended function(s). In addition, the NRC review should 

ensure that the response of the structures to credible off-normal and 

accident conditions will not create secondary hazards for cask system 

components or the stored nuclear materials. 

(2) Applicable Codes and Standards 

The reviewer should review the suitability of the applicant’s identification 

of codes and standards to be met by the structural design and 

construction of other components subject to NRC approval. The principal 

codes and standards include the following references although any of the 

previously identified codes in Sections 3.5.1.2.ii(2) and 3.5.2.2.i(2) may 

be used. 

 • ASCE 7. 

 • International Building Code (IBC). 

 • AISC, “Specification for Structural Steel Buildings—Allowable 

Stress Design and Plastic Design.” 

 • AISC, “Code of Standard Practice for Steel Buildings and 

Bridges.” 

 • ASME B&PV Code, Section VIII. 

 • ACI 318. 

 ii. Structural Design Features 

The reviewer should examine the adequacy of the applicant’s descriptions of 

cask system components that are not important to safety but are subject to NRC 

approval. These descriptions should adequately identify the intended function(s) 

of each component. 

Although the components evaluated in this portion of the DSS review are not 

directly important to safety, a credible possibility may exist that the structural 

response or failure of these components may cause a secondary risk to other 

components that are important to safety or to the subject nuclear material. For 

example, under tornado or seismic event conditions, the components may impact 
 3-39 

other components that are important to safety. When such a possibility exists, 

the applicant must provide more extensive structural information and greater 

assurance of acceptable fabrication and construction. 

3.5.3.3 Materials Related to Structural Evaluation 

The identification of structural materials should be reviewed in coordination with the materials 

discipline in Chapter 8 to the extent appropriate to determine if they are adequate for their 

intended function(s). The reviewer should determine the required level of review and extent of 

information in relation to the possibility and consequences of secondary effects on components 

that are important to safety. Materials should be as permitted or specified in the applicable 

code(s). 

3.5.3.4 Structural Analysis 

i. Load Conditions 

The load definitions and combinations shown in Tables 3-2 and 3-3 have been 

accepted by the NRC for analysis of steel and reinforced concrete ISFSI 

structures that are important to safety. These load combinations may also be 

used for structures not important to safety. 

In addition, for structures not important to safety, the NRC accepts the use of 

load combinations given in the IBC as well as ACI 349, ANSI/ANS 57.9, and 

ASCE 7. 

The NRC also accepts the load descriptions, combinations, and analytical 

approaches given in the ASME B&PV Code, Section VIII, for pressure systems, 

vessels, and casks that do not form elements of the confinement cask. 

ii. Structural Analysis Methods 

The reviewer should evaluate the applicant’s selection and use of structural 

analysis methods, codes, and models and ensure that these are consistent with 

and appropriate for the design code applicable to the component (as discussed 

above). 

iii. Structural Evaluation 

The reviewer may determine that an NRC structural evaluation of certain other 

components is not necessary for approval of the cask system. Similarly, the 

NRC may determine that approval of the cask system does not need to include 

specific components that are not important to safety, even though the applicant 

seeks approval of those components as part of the application. 

The SER should identify the system components that are excluded from the 

approval, stating the rationale for exclusion of each. As a corollary, the SER 

should also identify the components that are included, stating any limitations on 

the scope of the NRC review (e.g., “reviewed for functionality only”). 
 3-40 

3.6 Evaluation Findings 

The structural evaluation must provide reasonable assurance that the cask system will allow 

safe storage of SNF. This finding should be reached on the basis of a review that considered 

the regulation, appropriate RG, applicable codes and standards, and accepted engineering 

practices. Acceptance of the structural design of a storage cask system therefore implies that 

the design meets the relevant requirements of the following regulations: 

 F3.1 The SAR adequately describes all SSCs that are important to safety, providing 

drawings and text in sufficient detail to allow evaluation of their structural 

effectiveness. 

 F3.2 The applicant has met the requirements of 10 CFR Part 72.236(b). The SSCs 

important to safety are designed to accommodate the combined loads of normal 

or off-normal operating conditions and accidents or natural phenomena events 

with an adequate margin of safety. Stresses at various locations of the cask for 

various design loads are determined by analysis. Total stresses for the 

combined loads of normal, off-normal, accident, and natural phenomena events 

are acceptable and are found to be within limits of applicable codes, standards, 

and specifications. 

 F3.3 The applicant has met the requirements of 10 CFR Part 72.236(c), for 

maintaining subcritical conditions. The structural design and fabrication of the 

DSS includes structural margins of safety for those SSCs important to nuclear 

criticality safety. The applicant has demonstrated adequate structural safety for 

the handling, packaging, transfer, and storage under normal, off-normal, and 

accident conditions. 

 F3.4 The applicant has met the requirements of 10 CFR 72.236(l), “Specific 

Requirements for Spent Fuel Storage Cask Approval.” The design analysis and 

submitted bases for evaluation acceptably demonstrate that the cask and other 

systems important to safety will reasonably maintain confinement of radioactive 

material under normal, off-normal, and credible accident conditions. 

 F3.5 The applicant has met the requirements of 10 CFR 72.236 with regard to 

inclusion of the following provisions in the structural design: 

 - Design, Fabrication, Erection, and Testing to Acceptable Quality 

Standards. 

 - Adequate Structural Protection Against Environmental Conditions 

and Natural Phenomena, Fires, and Explosions. 

 - Appropriate Inspection, Maintenance, and Testing. 

 - Adequate Accessibility in Emergencies. 

 - A Confinement Barrier that Acceptably Protects the Cladding 

During Storage. 
 3-41 

 - Structures that are Compatible with Appropriate Monitoring 

Systems. 

 - Structural Designs that are Compatible with Retrievability of SNF. 

 F3.6 The Applicant has met the specific requirements of 10 CFR 72.236(g) and (h) as 

they apply to the structural design for spent fuel storage cask approval. The cask 

system structural design acceptably provides for the following required 

provisions: 

 - Storage of the Spent Fuel for a Minimum Required Years. 

 - Compatibility with Wet or Dry Loading and Unloading Facilities. 

The reviewer should provide a summary statement similar to the following: 

“The staff concludes that the structural properties of the structures, systems, and 

components of the [cask designation] are in compliance with 10 CFR Part 72, and that 

the applicable design and acceptance criteria have been satisfied. The evaluation of the 

structural properties provides reasonable assurance that the [cask designation] will allow 

safe storage of SNF for a licensed (certified) life of years. This finding is reached 

on the basis of a review that considered the regulation itself, appropriate regulatory 

guides, applicable codes and standards, and accepted engineering practices.” 

3.7 Designations and Descriptions of Loads 

Definitions of terms used in the following table are as accepted by the NRC. Many definitions 

are expanded with their intended applications more fully described and implemented than in the 

referenced sources. 

Tables 3-2 and 3-3 do not apply to the analysis of confinement casks and other components 

designed in accordance with Section III of the ASME B&PV Code. 

Capacities (“S” and “U” terms) and demands (factored or unfactored loads may be loads, forces, 

moments, or stresses caused by such loads. Usage must be consistent among the terms used 

in the load combination. Units of force, rather than mass, are to be used for loads. 

Definitions of terms used in the load combination expressions for reinforced concrete and steel 

are derived from ANSI 57.9, ACI 349, AISC specifications, or another source. Where used in an 

expression related to steel analysis, definitions derived from ACI 349 are not limited in 

application to reinforced concrete analyses. 

The load combinations defined on the basis of allowable stress apply to total stresses (that is, 

combined primary and secondary stresses). The load and stress factors do not change if 

secondary stresses are included. 
 3-42 

Table 3-2 Loads and Their Descriptions 

Symbol Capacity or Load Term Capacity or Load (or Demand) Description

S Steel ASD strength Strength of a steel section, member, or connection computed in 

accordance with the “allowable stress method” of the AISC 

“Specification for Structural Steel Buildings.” 

Sv Steel ASD shear strength Shear strength of a section, member, or connection computed in 

accordance with the “allowable stress method” of the AISC 

“Specification for Structural Steel Buildings.” 

Us Steel plastic strength Strength (capacity) of a steel section, member, or connection 

computed in accordance with the “plastic strength method” of the 

AISC “Specification for Structural Steel Buildings.” 

Uc reinforced concrete 

available strength 

Minimum available strength (capacity) of reinforced concrete 

section, member, or embedment to meet the load combination, 

calculated in accordance with the requirements and assumptions 

of ACI 349 and, after application of the strength reduction factor, 

Ø, as defined and prescribed at §9.2, “Design Strength,” of ACI 

349. If strength may be reduced during the design life by 

differential settlement, creep, or shrinkage, those effects shall be 

incorporated in the dead load, D (instead of by subtraction from 

minimum available strength) reinforced concrete footing and 

foundation sections whose demand loads are dominated by the 

maximum soil reaction may be designed and evaluated using Uf.

Uf Strength of foundation 

sections 

Minimum available strength of reinforced concrete footing and 

foundation sections whose demand loads are dominated by the 

maximum soil reaction, and after the strength reduction factor, 

Ø, as defined and prescribed at §9.3, “Design Strength,” of ACI 

349 is applied. Structural elements interface with columns, 

walls, grade beams, or footings and foundations should be 

evaluated by using load factors and load combinations for Uc. 

These interface elements include anchor bolts and other 

embedments, dowels, lugs, keys, and reinforcing extended into 

the footing or foundation. 

Ug Soil reaction or pile 

capacity 

Minimum available soil reaction or pile capacity is determined by 

foundation analysis (expressed in a SAR for approval of a cask 

system as a required minimum for the cask system design). 

Ug is derived using the same load factors and load combinations 

as shown for determination of Uc. 

O/S Overturning/ sliding 

resistance 

Required minimum available resistance capacity of structural 

unit against both overturning or sliding. Capacities for resistance 

of overturning and sliding are checked against the factored load 

combination separately, although the minimum margins of safety 

may occur concurrently. O/S is not determined by strength 

capacities of structural elements. Stress or strength demands 

resulting from an overturning or sliding situation are evaluated in 

load combinations involving S, Sv, Us, Uc, and Uf. 
 3-43 

Table 3-2 Loads and Their Descriptions 

Symbol Capacity or Load Term Capacity or Load (or Demand) Description

 All loads used in 

combination 

If any load reduces the effects of the combination of the other 

loads and that load would always be present in the condition of 

the specific load combination, the net coefficient (factor) for that 

load shall be taken as 0.90. If the load may not always be 

present, the coefficient for that load shall be taken as zero. 

Each load that may not always be present in the load 

combinations is to be varied from 0 to 100 percent to simulate 

the most adverse loading conditions (to the extent of proving that 

the lowest margins of safety have been determined). 

D Dead load Dead load of the structure and attachments including 

permanently installed equipment and piping. The weight and 

static pressure of stored fluids may be included as dead loads 

when these are accurately known or enveloped by conservative 

estimates. Loads resulting from differential settlement, creep, 

and/or shrinkage, if they produce the most adverse loading 

conditions, are included in dead load. If differential settlement, 

creep, or shrinkage would reduce the combined loads, it shall be 

neglected. D includes the weight of soil vertically over a footing 

or foundation for the purposes of determining Ug, Uf, and O/S. 

Regardless of the load combination factor applied, D is to be 

varied by +5 percent if that produces the most adverse loading 

condition. 

L Live loads Live loads, including equipment (such as a loaded storage cask) 

and piping not permanently installed, and all loads other than 

dead loads that might be experienced that are not separately 

identified and used in the load combination, and that are 

applicable to the situation addressed by the load combination. 

Typically includes the gravity and operational loads associated 

with handling equipment and routine snow, rain, ice, and wind 

loads, and normal and off-normal impacts of equipment. Loads 

attributable to piping and equipment reactions are included. 

Depending on the case being analyzed, may include normal or 

off-normal events not separately identified, as may be caused by 

handling (not including drop), equipment or instrument 

malfunction, negligence, and other man-made or natural causes. 

Live loads attributable to casks with stored fuel need only be 

varied by credible increments of loading of an individual cask. 

Live loads attributable to multiple casks should be varied for the 

presence and positioning of one or more cask(s), as necessary 

and varied to determine the lowest margins of safety. 
 3-44 

Table 3-2 Loads and Their Descriptions 

Symbol Capacity or Load Term Capacity or Load (or Demand) Description

L Live load for precast 

structures before final 

integration in-place 

Live loads for precast structures shall consider all loading and 

restraint conditions from initial fabrication to completion of the 

structure including form removal, storage, transportation, and 

erection. The NRC is concerned with analysis of loading of 

reinforced concrete structures before use to the extent that the 

structures should not have suffered hidden damage from 

construction live loads, thereby jeopardizing the capacity of the 

structures when in use. If the damage would be visibly obvious 

before installation, analysis of capacity versus pre-completion 

demands is not required. 

DB “Design-basis” (accidentlevel) loads Design-basis loads are controlling bounds for the following 

external event estimates: 

(1) Extreme credible natural events to be used for deriving 

design bases that consider historical data or rated 

parameters, physical data, or analysis of upper limits of 

the physical processes involved. 

(2) Extreme credible external man-induced events used for 

deriving design bases on the basis of analysis of 

human activity in the region taking into account the site 

characteristics and associated risks. 

Design-basis loads include credible accidents and extreme 

natural phenomena. Presumption of concurrent independent 

accidents or severe natural phenomena producing compounding 

design-basis loads is not required. Capacity to resist design 

basis loads can be assumed to be that of a structure that has not 

been degraded by previous design basis loads unless prior 

significant degradation in structural capacity may credibly occur 

and remain undetected. 

T Thermal loads Thermal loads, including loads associated with “normal” 

condition temperatures, temperature distributions, and thermal 

gradients within the structure; expansions and contractions of 

components; and restraints to expansions and contractions with 

the exception of thermal loads that are separately identified and 

used in the load combination. Thermal loads shall presume that 

all loaded fuel has the maximum thermal output allowed at time 

of initial loading in the cask system. Thermal loads shall be 

determined for the most severe of both steady-state and 

accident conditions. For multiple cask storage facilities, thermal 

loads shall be determined for the worst-case loadings on 

potentially critical sections (e.g., all in place, only one cask in 

place, alternating casks in place). 
 3-45 

Table 3-2 Loads and Their Descriptions 

Symbol Capacity or Load Term Capacity or Load (or Demand) Description

Ta Accident- level thermal 

loads 

Thermal loads produced directly or as a result of off-normal or 

design-basis accidents, fires, or natural phenomena. [Note: 

Although off-normal and design-basis thermal loads are treated 

the same in the load combinations, there is a distinction between 

off-normal and design-basis temperature limits for concrete. Offnormal temperature limits are the same as for “normal” 

conditions.] For multiple cask storage facilities, thermal loads 

shall be determined for the worst-case loadings on potentially 

critical sections. 

A Accident loads Loads attributable to the direct and secondary effects of an offnormal or design-basis accident as could result from an 

explosion, crash, drop, impact, collapse, gross negligence, or 

other man-induced occurrences; or from severe natural 

phenomena not separately defined. Loads attributable to direct 

and secondary effects may be assumed to be nonconcurrent 

unless they might be additive. The capacity for resistance to the 

demand resulting from secondary effects would be that residual 

capacity following any degradation caused by the direct effect. 

H Lateral soil pressure Loads caused by lateral soil pressure as would exist in normal, 

off-normal, or design-basis conditions corresponding to the load 

combination in which used. H includes lateral pressure resulting 

from ground water, the weight of the earth, and loads external to 

the structure transmitted to the structure by lateral earth 

pressure (not including earthquake loads, which are included in 

E, see below). H does not include soil reaction associated with 

attempted lateral movement of the structure or structural 

element in contact with the earth. 

G Loads attributable to soil 

reaction 

Used only in load combinations for footing and foundation 

structural sections for which demand is limited by the soil 

reactions. G represents loads attributable to the maximum soil 

reaction (horizontal (passive pressure limit) and vertical (soil or 

pile bearing limit) that would exist in normal, off-normal, or 

design-basis conditions corresponding to the load combination 

used. G is a function of Ug (i.e., G = f (Ug)). 

W Wind loads Wind loads produced by normal and off-normal maximum winds. 

Pressure resulting from wind and with consideration of wind 

velocity, structure configuration, location, height above ground, 

gusting, importance to safety, and elevation may be calculated 

as provided by ASCE 7. 
 3-46 

Table 3-2 Loads and Their Descriptions 

Symbol Capacity or Load Term Capacity or Load (or Demand) Description

Wt Tornado loads Loads attributable to wind pressure and wind-generated missiles 

caused by the design-basis tornado or design-basis wind (for 

sites where design-basis wind rather than tornado produces the 

most severe pressure and missile loads). Pressure resulting 

from wind velocity and elevation may be calculated as provided 

for these factors in ASCE 7. Tornado wind velocity or pressure 

does not have to be increased for structure importance, gusting, 

location, height above ground, or importance to safety (these do 

apply for design-basis wind). 

E Earthquake loads Loads attributable to the direct and secondary effects of the 

design earthquake or off-normal flood, including flooding caused 

by severe and extreme natural phenomena (e.g., seiches, 

tsunamis, storm surges), dam failure, fire suppression, and other 

accidents. 

3.7.1 Load Combinations for Steel and Reinforced Concrete Non-Confinement 

Structures 

The reinforced concrete structure load combinations apply to reinforced concrete structures 

important to safety that are not within the scope of ACI 359 (ASME B&PV Code, Section III, 

Division 2). The load combinations apply to steel structures important to safety that are not 

within the scope of the ASME B&PV Code, Section III, Division 1. The NRC accepts, but does 

not require use of these load combinations for steel and reinforced concrete structures that are 

not important to safety. The NRC accepts steel analyses that reflect allowable stress design or 

plastic strength design. Steel load combinations may be determined on the basis of the set of 

load combination expressions involving either “S” or “Us.” 

Table 3-3 Load Combinations for Steel and Reinforced Concrete Non-Confinement 

Structures 

Load Combination Acceptance Criteria

Reinforced Concrete Structures — Normal Events and Conditions

Uc > 1.4 D + 1.7 L Capacity/demand >1.00 for all sections. 

Uc > 1.4 D + 1.7 (L + H) Capacity/demand >1.00 for all sections. 

Reinforced Concrete Structures — Off-Normal Events and Conditions

Uc > 1.05 D + 1.275 (L + H + T) Capacity/demand >1.00 for all sections. 

Uc > 1.05 D + 1.275 (L + H + T + W) Capacity/demand >1.00 for all sections. 

Reinforced Concrete Structures — Accidents and Conditions

Uc > D + L + H + T + ( E or F) Capacity/demand >1.00 for all sections. 
 3-47 

Table 3-3 Load Combinations for Steel and Reinforced Concrete Non-Confinement 

Structures 

Load Combination Acceptance Criteria

Uc > D + L + H + T + A Capacity/demand >1.00 for all sections. An overturning 

accident for a cask in transfer or in separate storage on a pad 

is to be assumed unless more severe overturning also occurs 

as a result of a natural phenomenon. 

Uc > D + L + H + Ta Capacity/demand >1.00 for all sections. 

Uc > D + L + H + T + Wt The load combination (capacity/demand >1.00 for all sections) 

shall be satisfied without missile loadings. Missile loadings are 

additive (concurrent) to the loads caused by the wind pressure 

and other loads; however, local damage may be permitted at 

the area of impact if there will be no loss of intended function 

of any structure important to safety. 

Reinforced Concrete Footings/Foundations — Normal Events and Conditions

Uf > D + (L + G) Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Uf > D + (L + H+ G) Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Reinforced Concrete Footings/Foundations — Off-Normal Events and Conditions

Uf > D + (L + H + T + G) Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Uf > D + (L + H + T + W + G) Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Reinforced Concrete Footings/Foundations — Accident-Level Events and Conditions

Uf > D + L + H + T + E + G Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Uf > D + L + H + T + A + G Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Uf > D + L + H + Ta + G Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Uf > D + L + H + T + Wt + G Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Uf > D + L + H + T + F + G Capacity/demand >1.00 for all sections. For footing and 

foundation sections with load limited by soil reaction. 

Steel Structures Allowable Stress Design — Normal Events and Conditions

(S and Sv) > D + L Factored strength/demand >1.00 for all sections. 

(S and Sv) > D + L + H Factored strength /demand >1.00 for all sections. 
 3-48 

Table 3-3 Load Combinations for Steel and Reinforced Concrete Non-Confinement 

Structures 

Load Combination Acceptance Criteria

Steel Structures Allowable Stress Design — Off-Normal Events and Conditions

1.3 (S and Sv) > D + L + H + W Factored strength /demand >1.00 for all sections. 

1.5 S > D + L + H + T + W Factored strength/demand >1.00 for all sections. Thermal 

loads may be neglected when analysis shows that they are 

secondary and self-limiting in nature, and when the material is 

ductile. 

1.4 Sv > D + L + H + T + W Factored strength/demand >1.00 for all sections. Thermal 

loads may be neglected when analysis shows that they are 

secondary and self-limiting in nature, and when the material is 

ductile. 

Steel Structures Allowable Stress Design — Accidents and Conditions

1.6 S > D + L + H + T + 

(E or Wt or F) 

Factored strength/demand >1.00 for all sections. Thermal 

loads may be neglected when analysis shows that they are 

secondary and self-limiting in nature, and when the material is 

ductile. 

1.4 Sv > D + L + H + T + 

(E or Wt or F) 

Factored strength (allowable stress design)/demand >1.00 for 

all sections. Thermal loads may be neglected when analysis 

shows that they are secondary and self-limiting in nature, and 

when the material is ductile. 

1.7 S > D + L + H + T + A Factored strength/demand >1.00 for all sections. Thermal 

loads may be neglected when analysis shows that they are 

secondary and self-limiting in nature, and when the material is 

ductile. 

1.4 Sv > D + L + H + T + A Factored strength/demand >1.00 for all sections. Thermal 

loads may be neglected when analysis shows that they are 

secondary and self-limiting in nature, and when the material is 

ductile. 

1.7 S > D + L + H + Ta Factored strength/demand >1.00 for all sections. 

1.4 Sv > D + L + H + Ta Factored strength/demand >1.00 for all sections. 

Steel Structures Plastic Strength Design — Normal Events and Conditions

Us > 1.7 (D + L) Plastic capacity/demand >1.00 for all sections. 

Us > 1.7 (D + L + H) Plastic capacity/demand >1.00 for all sections. 

Steel Structures Plastic Strength Design — Off-Normal Events and Conditions

Us > 1.3 (D + L + H + W) Plastic capacity/demand >1.00 for all sections. 

Us > 1.3 (D + L + H + T + W) Plastic capacity/demand >1.00 for all sections. Thermal loads 

may be neglected when analysis shows that they are 

secondary and self-limiting in nature, and when the material is 

ductile. 
 3-49 

Table 3-3 Load Combinations for Steel and Reinforced Concrete Non-Confinement 

Structures 

Load Combination Acceptance Criteria

Steel Structures Plastic Strength Design — Accidents and Conditions

Us > 1.1 (D + L + H + T + 

(E or Wt or F)) 

Plastic capacity/demand >1.00 for all sections. Thermal loads 

may be neglected when analysis shows that they are 

secondary and self-limiting in nature, and when the material is 

ductile. The load combination (capacity/demand >1.00 for all 

sections) shall be satisfied without missile loadings. Missile 

loadings are additive (concurrent) to the loads caused by the 

wind pressure and other loads; however, local damage may be 

permitted at the area of impact if there will be no loss of 

intended function of any structure important to safety. 

Us > 1.1 (D + L + H + T + A) Plastic capacity/demand >1.00 for all sections. An overturning 

accident for a cask in transfer or in separate storage on a pad 

is to be assumed unless more severe overturning also occurs 

as a result of a natural phenomenon. Thermal loads may be 

neglected when analysis shows that they are secondary and 

self-limiting in nature, and when the material is ductile. 

Us > 1.1 (D + L + H + Ta) Plastic capacity/demand >1.00 for all sections. 

Overturning and Sliding — Normal and Off-Normal Events and Conditions

O/S ≥ 1.5 (D + H) Capacity/demand ≥1.00 for structure to be satisfied for both 

overturning and sliding. 

Overturning and Sliding — Accidents and Conditions

O/S ≥ 1.1 (D + H + E) Capacity/demand ≥1.00 for structure to be satisfied for both 

overturning and sliding. 

O/S ≥ 1.1 (D + H + Wt) Capacity/demand ≥1.00 for structure to be satisfied for both 

overturning and sliding. 
 3-50 

APPENDIX 3A - COMPUTATIONAL MODELING SOFTWARE 

Technical Review Guidance: 

Computational Modeling Software (CMS) Application

The staff does not endorse the use of any specific type or code vendor of CMS. Any 

appropriate CMS application could be used for analyses of cask or package components; 

however, for any CMS to demonstrate that a particular cask design satisfies regulatory 

requirements, adequate validation of that CMS must be demonstrated by the applicant. 

Descriptions of CMS validations can be contained within a given application or incorporated by 

reference. 

The reviewer should verify that the following information is provided in the SAR or related 

documentation (such as proprietary calculation packages or benchmark reports): 

 (1) details of the methodology used to assemble the computational models and the 

theoretical basis of the program used; 

 (2) a description of benchmarking against other codes or validation of the CMS 

against applicable published data or other technically qualified and relevant data 

that is appropriately documented; 

 (3) standardized verification problems analyzed using the CMS, including 

comparison of theoretically predicted results with the results of the CMS; and 

 (4) release version and applicable platforms. 

Once the information described above has been docketed, it need not be submitted with each 

subsequent application, but can be referred to in subsequent SARs or related documents. If an 

applicant changes their analysis methodology or changes the type or vendor of the CMS used, 

the applicant should submit either a revision of previously submitted information or include a 

clear explanation of the methodology changes, and their effects on the analysis in question, in 

subsequent SAR submittals. 

Modeling Techniques and Practices

Modeling techniques and practices used by applicants may need to be verified to demonstrate 

adequacy of the model. 

 • The reviewer should verify that the CMS and the options used by the applicant 

are appropriate for adequately capturing the behavior of a cask, package, or any 

components. 

Relevant input and results files or an equivalent detailed model description and output should be 

submitted with the original application. 

 • Analysis input files should be submitted in an electronic format that would most 

easily allow the solution to be executed by the staff, should the staff desire to do 

so. In-depth review of CMS models is most easily done with input files that 

contain individual commands used to develop the model and apply the various 
 3-51 

boundary conditions; therefore, a text input file format (versus database format) 

is preferred. 

 • Input files should be annotated in a way that clearly demonstrates the process 

behind building and solving models developed using CMS. A well annotated 

input file will expedite staff review and preclude the need for further clarification 

questions by the staff. 

 • Appropriate electronic media should be used for submitting case and support 

files. 

Computer Model Development

The reviewer should verify that the computer model used for the analysis is adequately 

described, either in the SAR or in other documentation, is geometrically representative of the 

cask design being analyzed, has addressed how material and manufacturing uncertainties 

might affect the analysis, has appropriate boundary conditions, and has no significant analysis 

errors. 

 • The reviewer should verify that the model description includes an adequate basis 

for the selection of parameters and/or components used in the analysis model 

(e.g., why was a particular element type applied in the analysis model?) 

 • The reviewer should verify that models sufficiently represent cask or package 

geometry and that adequate justification is provided for simplifications used. 

Models created with CMS are often simplified to reduce computer processing 

time. Models can often omit geometric details or use homogenized or smeared 

material properties to represent complex geometry or material combinations and 

still retain analytic accuracy. 

 • The reviewer should verify that the applicant has discussed how manufacturing 

and/or assembly tolerances and contact resistances will affect the analyses that 

have been conducted, if at all, in both the structural and thermal disciplines. The 

reviewer should also verify that the applicant has described how tolerances 

and/or contact resistances are accounted for, if applicable, in the cask or 

package analysis models that are submitted for review. 

 • The reviewer should verify that the applicant has provided a general discussion 

of how error, warning, or advisory messages generated by the software affect the 

analysis result (if applicable). When processing a computer model developed 

using CMS, the software will frequently provide error, warning, or advisory 

messages indicating a possible problem with the model that may or may not be 

sufficient to terminate processing. If the error/warning function has been 

disabled during processing, an explanation of why this is appropriate should be 

provided. 

 • The reviewer should verify that, within the specific disciplines, the dimensions 

and physical units used in the models developed are clearly labeled and mutually 

consistent. The fundamental units of time, mass, and length should be clearly 

identified. All other physical units derived must be consistent with the basic units 

adopted. For example, if the unit of length is the millimeter (mm), time in 
 3-52 

milliseconds (ms), and mass in gram (g), then, the mechanical force will have 

units of Newton (N), energy in milliJoule (mJ), and stress in megapascal (MPa). 

Verify that the input parameters are expressed in the units as assigned. If an 

applicant chooses not to adopt this uniformity of units, the appropriate conversion 

must be applied prior to processing input into CMS. Similar assurances must be 

provided for the output for the analysis solution. 

Computer Model Validation

 • The reviewer should verify that model validation done with applicable 

experiments or testing is properly documented and appropriate references are 

provided. 

 • The reviewer should ensure that if the applicant takes credit for modeling 

conservatisms, those conservatisms have been demonstrated through validation 

of the model or analysis methodology. For example, accounting for certain 

conditions that occur during the hypothetical accident condition (HAC) fire, such 

as combustion of materials, the turbulent flow of hot gasses in the pool fire 

environment, and material anomalies that may manifest themselves in a fire can 

be done with specialized CMS codes (specifically, coupled CFD-FEA codes such 

as Sandia National Lab’s CAFÉ code), high performance computer hardware and 

extended compute times. Each of these conditions can be treated in a 

conservative fashion using standard CMS; however, validation of the CMS 

against actual data (such as open pool fire test data or material combustion 

data), to demonstrate the applicability of the CMS under the HAC fire, for a 

configuration similar to that which is being modeled, would be necessary. 

Justification of Bounding Conditions/Scenario for Model Analysis

The applicant must determine the most damaging orientation and worst-case conditions for a 

given design and document how the analytic model was configured for the scenario. 

The reviewer should verify that the applicant provided sufficient justification for selecting the 

most damaging orientation and worst-case conditions. 

Description of Boundary Conditions and Assumptions

 • The reviewer should verify, as necessary, that boundary conditions and 

assumptions are addressed in the textual description included in the SAR or 

other documents (e.g., emissivity values, absorptivity values, convective 

coefficients, radiation view factors, symmetry planes, and rigid surfaces). This 

information should be presented in either tabular form or in a complete textual 

manner. Justifications and bases for such items should also be included in the 

textual description. 

 • Values or quantities indicating performance enhancements, i.e., increasing 

material conductivity values to mimic internal convection or substantially reduced 

design load factors (DLFs) reflecting an unusually high degree of impact 

damping, should be accompanied with justifications and should be closely 

reviewed and independently verified, if needed, by staff. 
 3-53 

Documentation of Material Properties

As needed, the reviewer should assess that: 

 (1) units for material properties are consistent throughout the individual SAR 

chapters. 

 (2) material properties for all applicable temperature ranges are included. 

 (3) references to materials used by the CMS application and specific material 

properties based on geometry (e.g., conductivity in the X, Y and Z directions), are 

listed in the SAR or related documents. 

Description of Model Assembly

 • The reviewer should verify that the types of elements used in the model are listed 

in the SAR, preferably in tabular format, along with the corresponding materials 

or components in which they are used in the analysis model. (i.e., the reviewer 

should quickly be able to discern what elements and materials are associated 

with specific components of the analysis model.) 

 • The reviewer should verify that a sufficient explanation of the logic behind the 

creation of each specific computer model is provided, for effective confirmatory 

calculations to be performed. 

 • The reviewer should verify that the applicant has provided annotated input files 

(as appendices to the SAR or in related documents), that clearly outline the 

various steps in building the computer models submitted. If input files are not 

provided or do not adequately describe model assembly, the applicant should 

provide an adequate explanation of how computer models were assembled using 

the CMS in the appropriate SAR chapters or related documents. 

Loads and Time Steps

 • The reviewer should verify that loads, load combinations, and, if used by the 

analytical code, the load steps utilized in the computer model are clearly 

explained by the applicant. The staff should evaluate all loads, how they are 

placed on the computer models, load combinations, and if used, the time steps 

applied in the analysis. 

 • The reviewer should verify that the time steps specified for the solution of the 

analysis are sufficiently small to accurately capture the behavior of the structures, 

systems, or components being modeled. 

 • The reviewer should verify that incremental time steps (or sub-steps) are 

adequately converged. Information of convergence may be obtained from the 

output generated by the execution of the analysis solution. 
 3-54 

Sensitivity Studies

The discussion of sensitivity studies should be included in the general Computer Model 

Development discussion, as noted above, with relevant references to examples included in the 

SAR or related documents. 

 • The reviewer should verify that the applicant has completed sensitivity studies for 

relevant CMS modeling parameters. This includes element type and mesh 

density, load step size, interfacing gaps or contact friction, material models and 

model parameters selection, and property interpolation, if applicable. For 

example, a mesh sensitivity study should be conducted not only for mesh density 

but also for mesh density/refinement in areas of thermal or structural concern or 

where performance of the material is crucial, such as seal areas, lid bolts, etc. A 

mesh sensitivity is also needed to make sure the analysis results are mesh 

independent. 

 • The reviewer should verify that the results of applicable sensitivity studies are 

clearly described in the SAR or related documentation and can be independently 

verified, if necessary. 

 • The reviewer should verify that the applicant’s documentation includes at least a 

brief discussion of the different models used in their mesh sensitivity studies. 

Results of the Analysis

 • The reviewer should verify that the SAR, or related document(s), include all 

relevant results (tabular and computer plots) for applicable load cases and load 

combinations evaluated for design code compliance, and that all governing 

results (stresses/deformation) are clearly identified in the tables and on plots. 

 • The reviewer should verify that results are consistent throughout the SAR, and 

that the correct results are used in calculations of other cask or package 

performance parameters (e.g., calculated temperatures used in the internal 

pressure calculation should be verified). 
 4-1 

4 THERMAL EVALUATION 

4.1 Review Objective 

The thermal review ensures that the cask and fuel material temperatures of the dry storage 

system (DSS) will remain within the allowable values or criteria for normal, off-normal, and 

accident conditions. This objective includes confirmation that the temperatures of the fuel 

cladding (fission product barrier) will be maintained throughout the storage period to protect the 

cladding against degradation that could lead to gross rupture. Also confirmed is the use by the 

applicant of acceptable analytical and/or testing methods in the Safety Analysis Report (SAR) 

when evaluating the DSS thermal design. 

4.2 Areas of Review 

As defined in Section 4.5, “Review Procedures,” a comprehensive thermal evaluation should 

encompass the following areas of review: 

Decay Heat Removal System

Material and Design Limits

Thermal Loads and Environmental Conditions

Analytical Methods, Models, and Calculations

 Configuration 

 Material Properties 

 Boundary Conditions 

 Computer Codes 

 Temperature Calculations 

 Pressure Analysis 

 Confirmatory Analysis 

4.3 Regulatory Requirements 

This section presents a summary matrix of the portions of the U.S. Code of Federal Regulations 

(CFR) Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, 

High-Level Radioactive Waste, and Greater Than Class C Waste,” Title 10, “Energy” (10 CFR 

Part 72) that are relevant to the review areas addressed by this chapter. The NRC staff 

reviewer should be familiar with the regulatory language in these sections. Table 4-1 matches 

the relevant regulatory requirements associated with this chapter to the areas of review. 
 4-2 

Table 4-1 Relationship of Regulations and Areas of Review 

Area of Review

10 CFR Part 72 Regulations

72.122 

(h)(1), (l) 

72.236 

(b), (f), (g), (h) 

Decay Heat Removal Systems ● ●

Material and Design Limits ●

Thermal Loads and Environmental 

Conditions ● ●

Analytical Methods, Models, and 

Calculations ● ●

4.4 Acceptance Criteria 

4.4.1 Decay Heat Removal System 

The applicant must provide a detailed description of the proposed cask heat removal system 

and its passive cooling characteristics. All major components are to be clearly identified and 

their contribution to heat-removal from the fuel thoroughly explained. The mechanism of heat 

removal (i.e., conduction, convection, radiation) for each component should also be discussed. 

Evidence must be provided by the applicant that the decay heat removal system will operate 

reliably under normal and loading conditions. 

All instrumentation used to monitor cask thermal performance should also be described. 

4.4.2 Material and Design Limits 

Cask components and fuel materials should be maintained between their minimum and 

maximum temperature limits for normal, loading, off-normal, and accident-level conditions to 

enable all components to perform their intended safety function. 

To guarantee cladding integrity of zirconium-based alloys, the maximum calculated fuel cladding 

temperature should not exceed 400°C (752°F) for normal conditions of storage and short-term 

loading operations, including cask drying and backfilling. A higher temperature limit may ONLY 

be used for low burnup spent nuclear fuel (SNF) (less than 45 GWd/MTU), as long as the 

applicant can demonstrate that the best estimate cladding hoop stress is equal to or less than 

90 MPa (13.1 ksi) for the temperature limit that is proposed. During loading operations, 

repeated thermal cycling should be limited to less than 10 cycles, with cladding temperature 

variations more than 65°C (149°F). For off-normal and accident conditions, the maximum 

zirconium based cladding temperature should not exceed 570°C (1058°F). 

To guarantee stainless steel cladding integrity, the maximum calculated fuel cladding 

temperature should not exceed 570°C (1058°F) for off-normal and accident conditions and the 
 4-3 

maximum calculated fuel cladding temperature should not exceed 400°C (752°F) for normal 

conditions of storage and short-term loading operations, including cask drying and backfilling. 

The applicant must clearly identify the operational temperature limits for all important-to-safety 

component materials under normal, loading, unloading, off-normal and accident-level 

conditions. The applicant shall provide reliable basis for all the temperature limits. 

The maximum internal pressure of the fuel container should remain within its design pressures 

for normal, off-normal, and accident-level conditions assuming rupture of 1 percent, 10 percent, 

and 100 percent of the fuel rods, respectively. Assumptions for pressure calculations include 

release of 100 percent of the initial fill gas and 30 percent of the fission product gases 

generated within the fuel rods during operation. 

The applicant must clearly identify the design pressure limits for the fuel container under normal, 

off-normal and accident-level conditions. 

4.4.3 Thermal Loads and Environmental Conditions 

Identification and justification of the design basis thermal load must be made by the applicant as 

well as the insolation and ambient temperature assumptions used as boundary conditions for 

the normal, loading, off-normal, and accident scenarios. 

4.4.4 Analytical Methods, Models, and Calculations 

The applicant shall present a thermal analysis that clearly demonstrates the storage system’s 

ability to manage design heat loads and have the various materials and components remain 

within temperature limits. The analysis shall be conducted for normal, loading, 

draindown/reflood, off-normal, and accident-level conditions. Resulting temperature profile and 

internal pressure information are necessary to support the structural analysis (Chapter 3) and 

the confinement analysis (Chapter 5) of the SAR. 

The applicant shall specify the analytical methods used in the thermal evaluations including any 

computational modeling software, (i.e., heat transfer or computational fluid dynamics computer 

analysis codes) and shall discuss the basis for the parameters and options selected for the 

analysis. All models should be clearly described. Material thermal properties for all cask 

components shall be provided and justified. The applicant must discuss, quantify, and report in 

the SAR any conservatism associated with the proposed thermal models. The level of detail of 

the discussion should be comparable with sections of the SAR that describes the analytical 

thermal models. A table of results should be provided in the SAR showing how the associated 

conservatisms affect the safety parameters (e.g. calculated peak cladding temperature, 

confinement seal temperatures, etc.). The table of results must be supported with fully 

documented analytical models and calculations. 

The computer codes used in the thermal evaluation should be well-verified and validated. The 

applicant must provide acceptable basis (e.g., benchmark efforts, published results) for the 

accuracy of the chosen computer code(s) and justification for its use in the proposed evaluation. 

A discussion of the resulting level of convergence and conservatism achieved as a function of 

the modeling options (e.g., meshing, time-differencing) must be provided by the applicant. 

To facilitate confirmatory analyses, electronic copies of the most significant input and output 

files should be provided. Further guidance on the review of analytical methods, models, and 
 4-4 

calculations provided to the staff for review is provided in Appendix 3A, “Computational 

Modeling Software.” 
 4-5 

Figure 4-1 presents an overview of the evaluation process and can be used as a guide to assist 

in coordinating with other review disciplines. 

Figure 4-1 Overview of the Thermal Evaluation 

Figure 4-1 Overview of the Thermal Evaluation

Chapter 1 – General 

Information Evaluation

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation

• DSS Description and 

Operational Features

• Drawings

• DSS Contents

• Thermal Tests

Chapter 12 – Accident 

Analysis Evaluation

• Model Specifications

• Temperatures

• Pressures

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation

• Cladding Heatup Limits

• Minimum Heat Load from 

Fuel

Chapter 3 – Structural 

Evaluation

• Temperatures

• Pressures

Chapter 9 – Operating 

Procedures Evaluation

• Subcooling Margin

• Cask Reflooding and 

Cooldown Quench 

Temperature and Flow

• Temperature Calculations

• Pressure Analysis

• Confirmatory Analysis

Analytical Methods, Models, and 

Calculations

• Configuration

• Material Properties

• Boundary Conditions

• Computer Codes

Decay Heat Removal System

Material and Design Limits

Thermal Loads and Environmental 

Conditions

Chapter 4 – Thermal Evaluation

• Temperature Calculations

• Pressure Analysis

• Confirmatory Analysis

Analytical Methods, Models, and 

Calculations

• Configuration

• Material Properties

• Boundary Conditions

• Computer Codes

Decay Heat Removal System

Material and Design Limits

Thermal Loads and Environmental 

Conditions

Chapter 4 – Thermal Evaluation

Chapter 2 – Principal 

Design Criteria Evaluation

• Spent Fuel Specifications

• General Design Criteria

• Thermal Design Criteria

Chapter 6 – Shielding 

Evaluation

• Radiation Source

• Gas Inventory

Chapter 8 – Materials 

Evaluation

• Material Properties

• Temperature Limit

Chapter 5 –

Confinement Evaluation

• Temperatures

• Pressures

Chapter 12 – Accident 

Analysis Evaluation

• Accident Conditions
 4-6 

4.5 Review Procedures 

Design features and acceptance criteria, initially presented in SAR Chapter 1, “General 

Information,” and Chapter 2, “Principal Design Criteria,” should be reviewed for additional insight 

about the thermal models that are being presented. Reviewers should examine the 

appropriateness of the proposed heat loads and environmental conditions. Modeling details 

such as simulation options, simplifications, and accuracy of results should be assessed. The 

DSS is to be analyzed under normal, loading, off-normal, and accident scenarios. If necessary, 

the resulting temperature distributions and internal pressures calculated in the SAR should be 

confirmed in order to verify compliance with design criteria and regulatory requirements. 

One of the most important results of the DSS thermal evaluation is confirmation that the fuel 

cladding temperature will remain below a specified limit to prevent unacceptable degradation 

during storage. 

Thermal performance of the cask under accident conditions is also evaluated in accordance 

with Chapter 12, “Accident Analyses Evaluation,” of this SRP, as appropriate, in the overall 

accident analyses presented in the SAR. 

In conducting a comprehensive thermal evaluation, reviewers should perform the established 

review procedures, as applicable, for each of the following areas of review. 

4.5.1 Decay Heat Removal System (HIGH Priority) 

The reviewer should examine the description of the DSS presented in SAR Chapter 1, “General 

Information Evaluation” as supplemented by the additional information provided in SAR Chapter 

4, “Thermal Evaluation.” These two sources of information should be consistent and 

supplementary. In addition to the material compositions, the dimensions of the cask 

components and SNF assemblies are to be clearly indicated. All drawings, figures, and tables 

should be sufficiently detailed to support in-depth staff evaluation. 

The applicant’s analysis should include the description of the significant thermal design features 

and operating characteristics of all pertinent DSS components and subsystems. Design 

features typically include the cask body, thermal fins, shielding materials, fuel baskets, heat 

transfer disks, confinement seals, drain and vent ports, and external pressure relief devices for 

the case of transfer casks, among others. The reviewer should verify that the thermal design 

features will adequately perform their intended safety functions during normal, loading, offnormal, and accident-level conditions. All thermal design features should be passive. 

Applicants have requested temporary external forced cooling of cask systems during loading 

operations or as a Technical Specification action statement during transfer operations. Such 

requests need to be examined by the staff to ensure that they meet the original intent of the 

regulations; that cask systems remain passively cooled during normal operations. 

Any instrumentation used to monitor cask thermal performance should also be described by the 

applicant in sufficient detail to support in-depth staff evaluation. The monitoring instrumentation 

components should have a safety classification (presented in SAR Chapter 2, “Principal Design 

Criteria Evaluation”) commensurate with their function and should be fully justified. Applicable 

operating controls and criteria, such as temperature criteria and surveillance requirements, 

should be clearly indicated in SAR Chapter 13, “Technical Specifications and Operational 

Controls and Limits” discussed in the Safety Evaluation Report (SER), and included in the 

Certificate of Compliance (CoC), as appropriate. 
 4-7 

4.5.2 Material and Design Limits (Priority - as indicated) 

(MEDIUM Priority) One of the most important results of the thermal evaluation is the 

confirmation that the fuel cladding temperature is sufficiently low to prevent cladding damage or 

potential failure during storage. Section 4.4.2, “Material and Design Limits,” of this SRP 

identifies the criteria for cladding temperature limits. The application must clearly agree with 

these criteria. 

(MEDIUM Priority) During licensing reviews, the thermal reviewer should ensure that either of 

the following criteria are used: (1) the maximum calculated temperatures for normal conditions 

of storage and for fuel loading operations do not exceed 400°C (752°F), or (2) the maximum 

calculated temperatures for normal conditions of storage do not exceed 400°C (752°) and that 

the materials reviewer has verified that the best estimate cladding hoop stress is less than 90 

MPa (13.1 ksi) for the maximum allowable temperature specified by the applicant for short-term 

fuel loading. If the applicants use the latter approach, the thermal reviewer will verify that the 

materials reviewer has verified that the cladding hoop stresses are less than 90 MPa (13.1 ksi) 

for each fuel assembly type (e.g., 14x14, 17x17, 9x9, etc.) proposed for storage. Cladding 

oxide thickness used to compute hoop stress should be evaluated by the materials reviewer. 

Since the hoop stress is dependent on the rod internal pressure, cladding geometry, and the 

temperature of the gases inside the rod, the staff will verify that the applicant has calculated the 

best estimate hoop stress corresponding to the rod internal pressure of the highest burnup fuel 

assemblies of the specific type of assembly. 

(MEDIUM Priority) To limit the amount of SNF that could be released from the cladding under 

off-normal conditions or accidents, the maximum calculated cladding temperatures should be 

maintained below 570°C (1058°F). 

(MEDIUM - bolted closure/LOW - welded closure) The reviewer should verify that temperature 

restrictions (upper and lower allowable limits) on all components important to safety (e.g., 

confinement, shielding, subcriticality, heat removal) during normal, loading, off-normal, and 

accident scenarios are clearly identified in the application and that the predicted thermal 

behavior of the entire DSS is indeed within the specified limits. The thermal reviewer should 

confirm with the assigned materials reviewer the acceptability of all proposed temperature limits. 

(LOW Priority) The maximum internal pressure of the fuel container should remain within its 

design limits for normal, off-normal, and accident-level conditions assuming rupture of 1 

percent, 10 percent, and 100 percent of the fuel rods, respectively. The thermal reviewer 

should confirm with the assigned structural reviewer the acceptability of the proposed design 

pressure limits. 

(HIGH Priority) Any operating scenario (loading or unloading) that results on a time-dependent 

limiting condition (e.g., number of hours allowed for vacuum drying before fuel cladding 

temperature reaches its allowable limit) should also be addressed in Chapter 13, “Technical 

Specifications and Operating Controls and Limits Evaluation,” of the SRP and should be 

included as a limiting condition for operation (e.g., technical specification) in the CoC, as 

appropriate. 
 4-8 

4.5.3 Thermal Loads and Environmental Conditions (Priority - as indicated) 

(LOW Priority) The reviewer should examine the specification for the design-basis fuel decay 

heat presented in SAR Chapter 2, “Principal Design Criteria Evaluation” and ensure that this 

decay heat is consistent with the specified fuel types, burnups, enrichments and cooling times, if 

included. Some applications, however, may provide a bounding decay heat load (kW/assembly) 

without specifying details about the SNF (design, enrichment, cooling time). 

(LOW Priority) The axial distribution for the decay heat sources should also be discussed by the 

applicant with clear justification for a bounding approach. The reviewer should expect a 

somewhat flat-at-the center axial distribution with a peak-to-average value in the range of 1.1 to 

1.2, tapering towards both ends. 

(MEDIUM Priority) In general, the NRC staff accepts insolation values presented in 10 CFR Part 

71 for 10 CFR Part 72 applications. Because of the large thermal inertia of a storage cask, the 

insolation values listed in 10 CFR Part 71.71 may be averaged over a 24-hour day assuming 

steady-state conditions. 

(MEDIUM Priority) The reviewer should verify that the ambient temperatures used for normal 

and off-normal condition evaluations do indeed bound the available historical temperature data 

for any suggested storage site (current or future). The National Oceanic Atmospheric 

Administration (NOAA) National Climatic Data Center provides temperature statistics for many 

American cities and regions. (http://www.ncdc.noaa.gov/oa/ncdc.html). 

(MEDIUM Priority) Loading and unloading evaluations should be established on the basis of the 

SNF pool’s technical specification maximum temperature limit (typically 46°C (115°F)). 

4.5.4 Analytical Methods, Models, and Calculations (MEDIUM Priority) 

For cask system components in which material properties and performance vary with 

temperature, the reviewer should examine the assumptions used in determining temperature 

maxima, minima, gradients, and differences for the cask system, as well as review the 

assumptions used to determine fuel cladding temperatures. The assumed temperature 

changes over time should result in the bounding conditions for the structural analysis. The 

calculated temperatures in the various cask system components should be compared to the 

limiting temperature criteria for the appropriate materials. Ferritic materials are subject to failure 

by brittle fracture at low temperatures. The reviewer should verify the assumed low 

temperatures for cask system handling operations for consistency with material properties. 

Ambient temperature restrictions may be appropriate for cask handling operations. Any limiting 

conditions regarding ambient temperatures should be addressed in SAR Chapter 13, as well as 

SER Chapter 13, “Technical Specifications and Operating Controls and Limits Evaluation,” and 

should be included as a limiting condition for operation (e.g., technical specification) in the CoC, 

as appropriate. 

Analysis for accident-level (“design-basis”) temperatures should not be considered to envelop 

the analysis of normal or off-normal temperatures. The acceptance criteria for normal and offnormal temperature demands for structural capacity will differ. Therefore, all three conditions 

should be analyzed. In addition, the duration over which accident temperature conditions may 

exist should be evaluated. 
 4-9 

4.5.4.1 Configuration (HIGH Priority) 

The reviewer should verify that any model used in the thermal evaluation is clearly described. 

Separate models and submodels may be used for the evaluation of different conditions (normal 

storage, loading, off-normal situations, and accidents). Coordination with the structural review is 

necessary to evaluate any damage that may result from accidents or natural phenomena 

events. All models should be shown as conservative. 

Examination by the reviewer of the sketches or figures of all models ensures their proper use in 

the thermal calculations and verifies that the dimensions and materials are consistent with those 

in the drawings of the actual cask, as presented in SAR Chapter 1, “General Information 

Evaluation”. If possible, the reviewer should examine the computer input files to verify 

consistency with the model sketches and engineering drawings. Differences between the actual 

cask configuration and the model should be identified, and the model should be shown to be 

conservative. 

Particular attention during the review should be paid to gaps between cask components. 

Tolerances should be considered so that the thermal resistance of each gap is treated 

conservatively. Gases (e.g., air, helium) assumed to be present in the gap shall be described 

and justified. If a specific gas other than air in the cask cavity or gaps between cask 

components is relied upon for heat removal, the reviewer should verify that the applicant shows 

that the gas is retained and that the gas is not diluted by other gases having lower thermal 

conductivities during the entire storage period. For cask components that are important to heat 

removal, manufacturing techniques for joining components, surface roughness, contact 

pressures, and gap conductance values should be adequately described and justified. 

The reviewer should verify that decay heat generated in the SNF is limited to the active fuel 

region of the assemblies. The model should specifically account for the peaking in the central 

region or provide another conservative approach. Heat from any other stored component (e.g., 

control rods), if applicable, should also be distributed appropriately. In addition, the positions of 

heat sources relative to other cask components should be identified. 

The application should address the thermal interaction among casks in an array by using a view 

factor less than unity. Generally, this will result in an operating control and limit in SAR Chapter 

13 that imposes a minimum spacing between storage casks. 

Coordination with the structural reviewer is necessary to ensure that the applicant has analyzed 

situations that may produce the worst-case cask loads. The greatest gradients and loadings 

caused by thermal expansion may occur with casks in alternative storage or in temporary 

handling positions. 

The heat transfer processes used in the analysis should be examined. Conduction and 

radiation are typically defined as the primary heat transfer mechanisms within the cask itself. In 

narrow regions of any orientation, little or no convective heat transfer will occur, and only 

conduction through the gas filled void spaces is assumed. Larger gas volume regions can 

experience a significant level of convective heat transfer. The staff suggests that the applicant 

demonstrate the existence of convection in the larger gas regions and quantify the contribution 

of convection heat transfer to the overall removal of heat from the package. Traditionally, the 

staff has maintained that natural convection in enclosed cavities should be validated through 

sufficient CFD calculations or physical experiments. 
 4-10 

4.5.4.1.1 General Guidance on Computational Fluid Dynamics Analyses (HIGH Priority) 

Since the computational resources necessary to fully resolve flow between individual fuel pins in 

a cask model with numerous fuel assemblies would be enormous, one acceptable approach 

would be to treat fuel assemblies as a porous media for applications seeking to credit heat 

removal from fuel via internal convection. The reviewer should verify that any CFD approach 

utilizes realistic or bounding flow friction factors in the porous media representation of the fuel, 

and that friction factors are obtained for each of the limiting fuel assembly types sought as 

approved contents for the cask. 

An acceptable approach to calculate the friction factors would be to perform a computational 

fluid dynamics (CFD) analysis for each type of fuel assembly for the expected operating 

conditions (pressure and average gas temperature). From the detailed CFD analysis of a single 

fuel assembly, wall shear stresses should be obtained separately for bare fuel rods and for fuel 

rods and associated grid straps. The friction factor shall be calculated based on the wall shear 

stress method. 

The reviewer should evaluate the method used to obtain the friction factors and ensure that the 

obtained values are realistic or bounding for the intended fuel assembly types. Also, since the 

friction factor is generally very sensitive to the geometric information (dimensions) and fuel 

assembly configuration, the reviewer should verify this information by reviewing the fuel 

assembly design drawings provided by the applicant. 

For ventilated spent fuel storage systems (a canister containing the fuel within an outer 

overpack), the mesh spacing (computational cell size) and density between an overpack liner 

and canister outer shell wall play an important role when selecting a turbulence model for the air 

flow through this annular gap. 

The near-wall modeling significantly impacts the fidelity of numerical solutions, inasmuch as 

walls are the main source of flow mean vorticity and turbulence. After all, it is in the near-wall 

region that the solution variables have large gradients, and the transport of momentum and 

other scalar variables occurs more vigorously. Therefore accurate representation of the flow in 

the near-wall region determines a successful prediction of wall-bounded turbulent flows. When 

dealing with wall effects on the flow usually two modeling options are available to the analyst. 

The first one is the use of the semi-empirical formulas called “standard wall functions” which are 

used to bridge the viscosity-affected region between the wall and the fully-turbulent core region. 

Generally a uniform mesh would be used when these wall functions are invoked. The use of 

wall functions obviates the need to modify the turbulence models to account for the presence of 

the wall. This modeling approach is usually applicable to flows with high Reynolds number. In 

the second approach, the viscosity-affected region is resolved with a mesh all the way to the 

wall, including the viscous sublayer. This type of approach is referred to as "near wall 

modeling" approach. The dimensionless distance between the wall and the cell center near the 

wall (y+) for the mesh used for this case should generally be around 1. Guidance on how to 

apply any of these modeling approaches should be provided in the CFD program 

documentation used in the application. Any modeling approach taken should be fully justified 

and validated. 

To properly characterize the flow (internal, external, annular, etc.), Reynolds number estimates 

shall be made using velocities from initial runs for the cooling air in the annulus and helium fill 

inside the canister. Reynolds number above 3000 based on the channel hydraulic diameter are 

above the critical Reynolds number of 2300 for internal flows, characterizing the flow in the 
 4-11 

transitional range between the laminar and turbulent zone. Since these are buoyancy driven 

flows, both the Grashof (Gr) number based on the hydraulic diameter of the channel and the 

modified Grashof number defined as Graetz number (Gz = Gr * W/H), where W and H are the 

width and height of the air channel,should also be calculated to properly characterize the 

annular flow. On the other hand, buoyancy driven helium flow, cooling the inside of the canister, 

generally would be laminar based on both the Grashof and the Reynolds numbers due to higher 

kinematic viscosities, and low achieved velocities within the canister. 

Actual SNF properties and uncertainties (e.g., friction factors, crud and oxide buildup, 

eccentricities, non-uniform axial and radial decay heat profiles) should also be addressed. 

Applicants must avoid using an effective thermal conductivity for the cover gas (e.g., helium) in 

lieu of a specific convection model. 

If applicable, the applicant should evaluate the added heat from components stored with the 

SNF assemblies (e.g., control rods, fuel channels, etc.). This would ultimately affect the 

maximum predicted cladding temperature. 

4.5.4.1.2 General Guidance on Application of Effective Conductivity Models (MEDIUM 

Priority) 

In addition to a CFD method utilizing a porous media, fuel assemblies may be modeled as a 

homogenous region using an effective thermal conductivity (this is a typical approach when 

utilizing a finite element analysis approach). The manner in which effective conductivity is 

determined for each fuel assembly should be examined by the reviewer. Guidance on effective 

thermal conductivity of the fuel is presented in Section 4.5.4.2, “Material Properties.” 

Use of effective thermal conductivity coefficients for regions within the confinement cask other 

than the fuel (e.g., gaps) may overestimate heat transfer. If effective thermal conductivity is 

used in this manner, the reviewer should verify that the same values have been determined 

from test data, or CFD submodels, or other appropriate sources that are representative of 

similar geometry, materials, temperatures, and heat fluxes used in current application. The 

reviewer should pay particular attention to the effective thermal conductivity of neutron shield 

regions, such as those embedded within thermal fins. Voids or gaps typically exist as a result of 

either tolerances or shrinkage, and should be considered in calculating effective thermal 

conductivity. Also, the applicant should pay particular attention to the values assumed for 

surface emissivities and view factors, as well as the manner used to account for radiation heat 

transfer in determining the effective thermal conductivities. 

4.5.4.2 Material Properties (MEDIUM Priority) 

The reviewer should coordinate with the materials discipline to verify that the material 

compositions and thermal properties are provided for all components used in the calculational 

model that the thermal properties used in the safety analysis are appropriate, and that potential 

degradation of materials over their service life has been evaluated. Temperature and 

anisotropic dependencies of thermal properties should be considered. If regional thermal 

properties are determined from a combination of individual materials, the manner in which these 

effective properties are calculated should be fully described and justified. 

If the thermal model is axisymmetric or three-dimensional, the longitudinal thermal conductivity 

should generally be limited to the conductivity of the cladding (weighted by its fractional area) 

within the fuel assembly. Gaps between fuel pellets and cracks in the pellets themselves can 
 4-12 

result in a considerable uncertainty regarding the contribution of the fuel to longitudinal heat 

transfer. High-burnup effects should also be considered in determining the fuel region effective 

thermal conductivity. 

4.5.4.3 Boundary Conditions (Priority - as indicated) 

(MEDIUM Priority) The reviewer should verify that the applicant identifies boundary conditions 

for normal, loading, off-normal, and accident conditions. The required boundary conditions 

include the decay heat rate from each fuel assembly and the external conditions on the cask 

surface. The peak power factor for a fuel assembly should be specified and the peak linear 

power (“peaking factor”) of a fuel assembly should be stated for a given active fuel length. 

(MEDIUM Priority) The boundary conditions on the cask surface depend on the environment 

surrounding the cask. Consequently, the temperature of the environment should be specified 

for all simulated conditions, as should the incident and absorbed insolation. The mechanisms 

and models for dissipating the absorbed insolation and decay heat from the surface of the cask 

to the environment should also be identified and described. The mechanisms for transferring 

heat from the cask surface usually consist of natural (free) convection and thermal radiation. A 

heat balance on the surface of the cask should be conducted and the results presented in the 

applicant’s SAR. 

(LOW Priority) The initial temperature distribution of the storage cask system before a fire 

accident should be established on the basis of the hottest temperature distribution during 

normal or off-normal storage conditions. The duration and flame temperature of the fire should 

be specified, as should gas velocities and flame emissivity. The flame and cask surface 

emissivities specified in 10 CFR 71.73(c)(4) for a hypothetical accident test of transportation 

packages are satisfactory for use with regard to a fire accident involving a storage cask. 

(LOW Priority) The applicant should identify and describe the mechanisms and models for 

coupling the fire energy to the cask surface. These mechanisms include forced convection in 

relation to the flame velocity (5 to 15 m/s, or about 16 to 49 ft/s) as well as thermal radiation. In 

addition, justification of the convection coefficients during the fire should be provided. Natural 

convection coefficients are not appropriate; as such coefficients imply downward gas flow 

adjacent to relatively cool cask walls. In general, for the fire condition, buoyant, upward flow, 

driven by hot gasses, will dominate. The orientation of the cask should also be considered. 

(LOW Priority) Following the fire, the cask is subject to insolation and content decay heat while 

being cooled by natural convection and thermal radiation to the environment. The applicant 

should identify the post-fire conditions of the cask, including any changes in surface conditions 

and/or geometry that may affect radiation and convection heat losses. Identification and 

description of the models used for the analysis of the post-fire processes should also be 

provided by the applicant. 

4.5.4.4 Computer Codes (HIGH Priority) 

The reviewer should verify that the applicant has provided information on any computer-based 

modeling as described in Appendix A to Chapter 3.0, “Structural Evaluation,” and review the 

thermal analysis submitted by the applicant in accordance with the Appendix. 
 4-13 

4.5.4.5 Temperature Calculations (Priority – as indicated) 

(MEDIUM - bolted closure/LOW - welded closure) The application should include a table that 

lists the maximum and minimum temperatures of all components important to safety under 

normal, loading, off-normal, and accident-level conditions. This table should specify the 

operating temperature range for each component. The reviewer should verify that temperatures 

have been calculated for key components and that they do not exceed the allowable range for 

each. Justification shall be provided in the application for any material important to safety that 

exceeds acceptable temperature ranges. If compliance with minimum temperature criteria 

relies on a specific minimum heat load from the fuel, such heat load shall be quantified and 

included as an operating control and a technical specification criterion in SAR Chapter 13. 

(MEDIUM Priority) The reviewer should pay particular attention to the maximum temperature of 

the cladding. These temperature limits are discussed in Section 4.4.2, “Material and Design 

Limits,” with review guidance presented in Section 4.5.2, “Material and Design Limits.” 

(MEDIUM Priority) Some storage systems rely upon natural circulation of air through internal 

passages to remove heat from the stored confinement canister. For storage systems with 

internal air flow passages, blockage of inlet and/or outlet flow is an accident situation that should 

be evaluated. Total blockage of all inlets and outlets may result in fuel heatup, which has been 

assumed to approach adiabatic conditions. To ensure that blockages do not go undetected for 

significant periods, the NRC has required objective evidence that inlet and outlet flows are not 

obstructed. Consequently, for these types of storage systems, the NRC has accepted periodic 

visual inspection of the vents coupled with temperature measurements to verify proper thermal 

performance and detect flow blockages. The inspections should take place within an interval 

that will allow sufficient time for corrective actions to be taken before the accident temperature is 

reached. The inspection interval should be more frequent than the time interval required for the 

fuel to heatup to the established accident temperature criteria, assuming a total blockage of all 

inlets and outlets. 

(MEDIUM Priority) The review of the heatup calculations should specifically address any 

assumptions regarding limiting components and quasi-steady state responses. The initial 

ambient temperature for the heatup calculations should bound the maximum “normal condition” 

temperature. The resulting heatup time history should be included in the SAR documentation, 

and should support the proposed inspection and monitoring intervals. This information is also 

useful in developing contingency operation procedures, since it indicates the available time in 

which to take corrective actions before the fuel accident temperature criteria may be exceeded. 

(HIGH Priority) Some storage systems may use a transfer cask to move the loaded confinement 

canister from the fuel handling building to the independent spent fuel storage installation (ISFSI) 

site. When the canister is within the transfer cask, the rate of cooling is typically less than for 

normal operation. Therefore, fuel cladding temperatures are expected to be higher than for 

normal storage conditions. 

(HIGH Priority) The reviewer should examine the temperature distribution calculations for the 

canister inside the transfer cask and verify that heat transfer through gap regions has been 

treated in a conservative manner, and that material properties and dimensions of the transfer 

cask are consistent with the design data defined in the SAR documentation. The initial ambient 

temperature should be the maximum “normal condition” temperature. Cask preparation for 

storage or unloading operations may include situations in which the canister is evacuated while 

it is in the transfer cask. If the fuel cladding temperature calculation is based on heatup over a 
 4-14 

limited time period for cask drying operations, the reviewer should verify that limiting conditions 

for the operations have been imposed in the technical specifications. Such limiting conditions 

should ensure that the temperature will remain acceptable during the operations, and that 

normal cooling will begin before the temperature criterion is exceeded. 

(HIGH Priority) During wet fuel transfer operations, the liquid in the fuel canister should not be 

permitted to boil. This practice avoids uncontrolled pressures on the canister and the connected 

dewatering, purging, and recharging system(s), unacceptable discharge of liquids which may be 

providing radiation shielding, and a potentially unacceptable reduction in the safety margin. The 

reviewer should ensure that to prevent any of the above conditions, an adequate subcooling 

margin is identified in both the SAR and corresponding operating procedures to prevent boiling. 

This margin may be cask-specific, depending on the design of the fuel basket and key 

assumptions used in the criticality analysis. The reviewer should ensure that the applicant 

reviews the heatup and time-to-boil calculations and assesses whether any technical 

specification or limiting conditions for operation are needed. Heatup calculations should be 

established on the basis of the SNF pool’s technical specification maximum temperature limit 

(typically 46°C (115°F)). 

(HIGH Priority) For unloading operations, the thermal reviewer should ensure that the applicant 

evaluates temperature and pressure calculations supporting procedural steps presented in SAR 

Chapter 9, “Operating Procedures Evaluation,” for cask cooldown and reflooding of the cask 

internals. To ensure that the cask does not overpressurize and that the fuel assemblies are not 

subjected to excess thermal stresses, the applicant’s analysis should specify and justify the 

appropriate temperature and flow rate of the quench fluid, assuming maximum fuel cladding 

temperatures in the unloading configuration. This analysis should also be referenced in Chapter 

12, “Accident Analyses Evaluation,” of the SAR as having been considered in the development 

of thermal models for the unloading procedures, and be included, as appropriate, in the 

technical specifications The thermal reviewer should provide thermal profiles to the materials 

reviewer so that latter can determine if the applicant has adequately addressed the issue of fuel 

rod response to a reflood incident in Chapter 8, “Materials Evaluation”. 

(LOW Priority) The most extreme thermal conditions may result from credible ambient 

temperatures, temperature-time histories, an adjacent fire, or any off-normal or design-basis 

event (DBE) resulting in blockage of ventilation passages. The worst-case structural loads may 

occur at temperatures lower than those of design-basis accidents (DBAs) or natural phenomena 

since load combination expressions effectively require greater safety factors for normal and offnormal analyses than for any DBE. Typically, fire has been the worst-case accident thermal 

condition for storage systems without internal air flow passages. 

(LOW Priority) The burning of fuel and other combustibles associated with vehicles involved in 

transfer operations should, at a minimum, be presumed to be a DBE with the cask in the most 

exposed situation during transfer or loading into storage. Fire parameters included in 10 CFR 

71.73 have been accepted for characterizing the heat transfer during the in-storage fire. 

However, a bounding analysis that limits the fuel source thus limits the length of the fire (e.g., by 

limiting the source to the fuel in the transporter) has also been accepted. 

(LOW Priority) Some structures, systems, and components (SSC) may experience the most 

severe conditions if exposure to high temperatures is followed by dousing with water (such as 

rain or fire suppression activities). A small amount of exterior concrete spalling may result from 

a fire, the application of fire suppression water, rain on heated surfaces or other hightemperature condition. The damage from these events is readily detectable and appropriate 
 4-15 

recovery or corrective measures may be presumed. Therefore, the loss of such a small amount 

of shielding material is not expected to cause a storage system to exceed the regulatory 

requirements in 10 CFR 72.106 and need not be estimated or evaluated in the SAR. The NRC 

accepts that concrete temperatures may exceed the temperature criteria of American Concrete 

Institute (ACI) 349 for accidents if the temperatures result from a fire. In that case, corrective 

action may be required for continued safe storage. 

(LOW Priority) The methods that are acceptable for analyzing and reviewing the consequences 

of a fire depend upon the duration of the fire and the margin between the predicted 

temperatures and the actual thermal limits of the components. A fire of sufficient duration, or 

one in which material temperatures are close to the criteria of their acceptable operational 

range, will require a detailed model of the cask and its contents. Cask system components 

(e.g., the neutron shield) may be assumed to be intact at the start of the fire. 

(LOW Priority) If a cask tipover is a credible accident, the reviewer should verify that the 

applicant has evaluated the effect on cask and fuel temperatures in the new configuration. An 

analysis may be warranted when a significant portion of heat removal capability is attributed to 

internal convection if a change in orientation of that cask may have a significant effect. 

4.5.4.6 Pressure Analysis (LOW Priority) 

Pressure calculations should be performed using the ideal gas law (i.e., PV = nRT where P is 

pressure, V is volume, n is the number of moles of a gas, R is a constant for a given gas, and T 

is the absolute temperature) and summing the partial pressures of each of the gas constituents 

in the cask cavity. The application should identify the method and all assumptions used in the 

pressure analysis, including the determination of the fission gas inventory. 

It is necessary to consider the temperature distribution of all components within the cask cavity 

and the cavity walls in calculating the gas pressure in the cavity. For the fire accident analysis, 

the application should identify the maximum gas temperature reached during the post-fire 

accident phase, explain the method used to determine the average gas temperature, and 

specify the time in the accident at which the peak gas temperature is attained. 

This pressure also depends on the free volume in the cask cavity, the amount (moles) of cover 

gas (helium) in the cavity, and the amount of gases released from ruptured fuel pins. The free 

volume calculation should be reviewed to determine if all components internal to the cask cavity 

(e.g., fuel assemblies, basket, structural supports, spacer disks, reactor control components) 

have been properly considered. 

The NRC accepts that normal conditions occur with less than 1 percent of the fuel rods failed, 

off-normal conditions occur with up to 10 percent of the fuel rods ruptured, and 100 percent of 

the fuel rods will have ruptured following a DBE. The NRC also accepts that a minimum of 

100 percent of the fill gas and 30 percent of the significant radioactive gases (e.g., 3

H, Kr, and 

Xe) within a ruptured fuel rod is available for release into the cask cavity. 

Under the conditions where any of the cask component temperatures are close (within 5 

percent) to their limiting values during an accident or the Maximum Normal Operating Pressure 

(MNOP) is within 10 percent of its design basis pressure, or any other special conditions, the 

applicant should consider, by analysis, the potential impact of the fission gas in the canister to 

the cask component temperature limits and the cask internal pressurization. 
 4-16 

The reviewer should coordinate with the structural reviewer to verify that the confinement 

pressure of the cask is within its design limits for normal and accident conditions. 

4.5.4.7 Confirmatory Analysis (HIGH Priority) 

Reviewers may need to perform a confirmatory analysis of the thermal performance of the cask 

SSCs identified as important to safety. Confirmatory analyses are recommended where 

margins between the calculated temperatures and prescribed component temperature limits are 

small, where particularly complex thermal analyses are submitted by applicants, or where the 

applicant is submitting a new thermal methodology or analysis approach. 

The application should be reviewed to ensure that the applicant made the correct assumptions 

and provided the correct input, and that the output is consistent with established physical 

(thermal) behavior. These results should specifically include steady-state temperature 

distributions, local heat balances, temperatures reached and temperature distributions within 

any reinforced concrete SSCs, and cask cavity pressures for the bounding ambient 

temperatures. 

To provide the most reliable confirmation, confirmatory analysis should, to the degree possible, 

use a different thermal analysis method than that used by the applicant. The code used for the 

confirmatory analysis may be the same as or different from that used by the applicant. 

Regardless, a review of the applicant’s analytical approach and analysis models should be 

considered part of the overall confirmatory analysis. Similar confirmation of accident 

temperatures (e.g., during a fire) should be performed, as applicable to the SAR analysis. 

If a full confirmatory analysis is not deemed necessary, the minimum confirmatory review should 

include verifying that key design parameters have been appropriately determined and correctly 

expressed as input into the computer program(s) used for the thermal analysis. Key parameters 

include proper dimensions, material properties (including surface emissivities and view factors 

for radiation), and definition of heat sources. A heat balance at the outer surface of the cask 

should be performed to verify that the heat from the SNF and insolance, balance that removed 

by convection and radiation. Correlations for the heat transfer coefficient should then be 

assessed to confirm that they are appropriate for the existing storage conditions. The 

temperature of the cask’s inner surface should be estimated by calculating the temperature 

distribution across the cask body with simple heat balance approximations. Finally, the 

difference between the cask’s inner surface temperature and the maximum cladding 

temperature should be compared with that of similar casks and baskets reviewed in previous 

SARs. 

As discussed above, a more detailed confirmatory analysis may be required, and could include 

a model of a portion of the cask or basket to ensure that the SAR results are realistic and 

conservative. A more extensive confirmatory analysis may involve the full geometry of the cask, 

with relevant component details, to determine temperature distributions in the cask system. 

Additional guidance on review of analytical models and conduct of confirmatory analyses can be 

found in Appendix 3A, “Computational Modeling Software.” 

As an alternative to a confirmatory analysis, the applicant may be required to perform designverification testing of an as-built cask or properly scaled mock-up system (when applicable) to 

confirm the thermal analyses presented in the SAR. Such testing may include verifying gap 

conductance values assumed in modeling thermal resistance. The test conditions, 
 4-17 

configuration, and type and location of instrumentation used, if any, should be sufficiently 

described in SAR Chapter 10, “Acceptance Criteria and Maintenance.” 

4.6 Evaluation Findings 

The reviewer should review the 10 CFR Part 72 acceptance criteria and provide a summary 

statement for each. These statements should be similar to the following model: 

 F4.1 Structures, systems, and components (SSCs) important to safety are described 

in sufficient detail in Chapters of the SAR to enable an evaluation of their 

thermal effectiveness. Cask SSCs important to safety remain within their 

operating temperature ranges. 

 F4.2 The [cask designation] is designed with a heat-removal capability having 

verifiability and reliability consistent with its importance to safety. The cask is 

designed to provide adequate heat removal capacity without active cooling 

systems. 

 F4.3 The spent fuel cladding is protected against degradation leading to gross 

ruptures by maintaining the cladding temperature for -year cooled fuel 

below °C (___ °F) in an [applicable gas] environment. Protection of the 

cladding against degradation is expected to allow ready retrieval of spent fuel for 

further processing or disposal. 

The reviewer should provide a summary statement similar to the following: 

“The staff concludes that the thermal design of the [cask designation] is in compliance 

with 10 CFR Part 72, and that the applicable design and acceptance criteria have been 

satisfied. The evaluation of the thermal design provides reasonable assurance that the 

[cask designation] will allow safe storage of spent fuel for a licensed (certified) life of 

years. This finding is reached on the basis of a review that considered the regulation 

itself, appropriate regulatory guides, applicable codes and standards, and accepted 

engineering practices.” 
 5-1 

5 CONFINEMENT EVALUATION 

5.1 Review Objective 

In this portion of the dry storage system (DSS) review, the U.S. Nuclear Regulatory Commission 

(NRC) evaluates the confinement features and capabilities of the proposed cask system. In 

conducting this evaluation, the NRC staff seeks to ensure that radiological releases to the 

environment will be within the limits established by the regulations and that the spent fuel 

cladding and fuel assemblies will be sufficiently protected during storage against degradation 

that might otherwise lead to gross ruptures. 

5.2 Areas of Review 

This chapter of the DSS Standard Review Plan (SRP) provides guidance for use in evaluating 

the design and analysis of the proposed cask confinement system for normal, off-normal, and 

accident conditions. This evaluation includes a more detailed assessment of the confinementrelated design features and criteria initially presented in Chapters 1, “General Information 

Evaluation” and 2, “Principal Design Criteria Evaluation” of the applicant’s Safety Analysis 

Report (SAR), as well as the proposed confinement monitoring capability, if applicable. In 

addition, the NRC staff assesses the potential releases of radionuclides associated with spent 

fuel by independently estimating their potential leakage to the environment and the subsequent 

impact on a hypothetical individual located at or beyond the controlled area boundary. 

As prescribed in U.S. Code of Federal Regulations (CFR) Part 72, “Licensing Requirements for 

the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste,” Title 10, 

“Energy” (10 CFR Part 72), the regulatory requirements for doses at and beyond the controlled 

area boundary include both the direct dose and that from an estimated release of radionuclides 

to the atmosphere (based on the tested leak tightness of the confinement). Thus, an overall 

assessment of the compliance of the proposed DSS with these regulatory limits is deferred to 

Chapter 11, “Radiation Protection Evaluation,” of this SRP. In addition, the performance of the 

cask confinement system under accident-level conditions, as evaluated in this chapter, may also 

be addressed in the overall accident analyses as discussed in Chapter 12, “Accident Analyses 

Evaluation,” of this SRP. 

As described in SRP Section 5.5, “Review Procedures,” a comprehensive confinement 

evaluation should encompass the following areas of review: 

Confinement Design Characteristics

Design Criteria 

Design Features 

Confinement Monitoring Capability 

 Nuclides with Potential for Release 

 Confinement Analyses

Normal Conditions 

Off-Normal Conditions (Anticipated Occurrences) 

Design Basis Accident Conditions (Including Natural Phenomenon Events) 

Supplemental Information
 5-2 

5.3 Regulatory Requirements 

This section presents a summary matrix of the portions of 10 CFR Part 72 that are relevant to 

the review areas addressed by this chapter. The NRC staff reviewer should read the exact 

referenced regulatory language. Table 5-1 matches the relevant regulatory requirements 

associated with this chapter to the areas of review. 

Table 5-1 Relationship of Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Regulations

72.104 (a) 72.122(a), (b)(1), 

(h)(1), (4), (i) 

72.236 

(d), (e), (i), (j), (l) 

Confinement Design Characteristics ● ●

Confinement Monitoring Capability ● 

Nuclides with Potential for Release ● ●

Confinement Analyses ● ● ●

5.4 Acceptance Criteria 

In general, the DSS confinement evaluation seeks to ensure that the proposed design fulfills the 

following acceptance criteria that the NRC staff considers to be minimally acceptable to meet 

the confinement requirements of 10 CFR Part 72. 

5.4.1 Confinement Design Characteristics 

The design should provide redundant sealing of the confinement boundary (10 CFR 72.236(e)). 

Typically, this means that field closures of the confinement boundary should either have two 

seal welds or two metallic O-ring seals. 

The confinement design should be consistent with the regulatory requirements as well as the 

applicant’s “General Design Criteria” reviewed in Chapter 2, “Principal Design Criteria 

Evaluation,” of this SRP. The NRC staff has previously accepted construction of the primary 

confinement barrier in conformance with the American Society of Mechanical Engineers (ASME) 

Boiler and Pressure Vessel (B&PV) Code, Section III, “Rules for Construction of Nuclear Facility 

Components,” Division 1, Subsections NB or NC. This code defines the standards for all 

aspects of construction including materials, design, fabrication, examination, testing, inspection, 

and certification required in the manufacture and installation of components. In such instances, 

the staff has relied upon Section III to define the minimum acceptable margin of safety. 

Therefore, the applicant must fully document and completely justify any deviations from the 

specifications of Section III. In some cases, after careful and deliberate consideration, the staff 

has made exceptions to this requirement. In addition, the ASME has published in 2005 Division 
 5-3 

3 to Section III which is written specifically for Containments for the Transportation and Storage 

of Spent Nuclear Fuel and is considered to be the governing code for this component, but has 

not yet been reviewed and endorsed by the NRC. 

The design must provide a nonreactive environment to protect fuel assemblies against fuel 

cladding degradation, which might otherwise lead to gross rupture (PNL, 1987). Measures for 

providing a nonreactive environment within the confinement cask typically include drying and 

backfilling with a nonreactive cover gas (such as helium). Experimental data have not 

demonstrated an acceptably low oxidation rate for UO2 spent fuel over the 20-year licensing 

period to permit safe storage in an air atmosphere during dry storage. Therefore, to reduce the 

potential for fuel oxidation and subsequent cladding failure, an inert atmosphere (e.g., helium 

cover gas) has been used for storing UO2 spent fuel in a dry environment. See Chapter 9, 

“Operating Procedures Evaluation,” of this SRP for more detailed information on the cover gas 

filling process. Note that other fuel types, such as graphite fuels for the high-temperature gascooled reactors (HTGRs), may not exhibit the same oxidation reactions as UO2 fuels and, 

therefore, may not require an inert atmosphere. Applicants proposing to use atmospheres other 

than inert gas should discuss how the fuel and cladding will be protected from oxidation. 

5.4.2 Confinement Monitoring Capability 

The reviewer should ensure the application describes the proposed monitoring capability and/or 

surveillance plans for mechanical closure seals. In instances involving welded closures, the 

staff has previously accepted that no closure monitoring system is required. This practice is 

consistent with the fact that other welded joints in the confinement system are not monitored, 

since the initial staff review ensures the integrity of the confinement boundary for the licensing 

period. Typical surveillances include checking for blockage of the air vents or temperature 

monitoring. 

To show compliance with the requirement for continuous monitoring, 10 CFR Part 72.122(h)(4), 

cask vendors have proposed, and the staff has accepted, routine surveillance programs and 

active instrumentation to meet the continuous monitoring requirements. 

5.4.3 Nuclides with Potential for Release 

The applicant must estimate the maximum credible quantity of radionuclides with potential for 

release to the environment. The radionuclides potentially available for release to the 

environment are based on the radiological source term evaluation presented in Chapter 6, 

“Shielding Evaluation,” of this SRP. 

5.4.4 Confinement Analyses 

The application should specify the maximum allowed leakage rates for the total primary 

confinement boundary and redundant seals. Applicants frequently display this information in 

tabular form including the leakage rate of each seal. The maximum allowed leakage rate is the 

“as tested” leak rate measured by the leak test performed on the cask field closure. Generally, 

as discussed below, the allowable leakage rate must be evaluated for its radiological 

consequences and its effect on maintaining an inert atmosphere within the cask. However, the 

analyses discussed below are unnecessary1

 for storage casks including its closure lid that are 

 

1

 For casks that are demonstrated to be leak tight, the review procedures discussed in Sections 5.5.3 and 5.5.4 are 

not applicable. 
 5-4 

designed and tested to be “leak tight” as defined in the American National Standards Institute 

(ANSI), Institute for Nuclear Materials Management’s “American National Standard for Leakage 

Tests on Packages for Shipment of Radioactive Materials” (ANSI N14.5-1997). 

 • The analysis of potential releases should be consistent with the methods 

described in ANSI N14.5-1997 (ANSI, 1997). 

 • During normal operations and anticipated occurrences, dose calculations based 

on the allowable leakage rate must demonstrate that the annual dose equivalent 

to any real individual who is located beyond the controlled area does not exceed 

the limits given in 10 CFR 72.104(a). 

 • For any design-basis accident, dose calculations based on the allowable leakage 

rate must demonstrate that an individual at the boundary or beyond the nearest 

boundary of the controlled area does not receive a dose that exceeds the limits 

given in 10 CFR 72.106(b)-(discussed further in Chapter 12, “Accident Analyses 

Evaluation”) 

 • The analysis of potential releases must demonstrate that an inert atmosphere will 

be maintained within the cask during the storage lifetime. 

 • For casks that employ a pressurized inert gas to facilitate internal natural 

convection heat transfer, the analysis of potential releases must demonstrate that 

the pressurized atmosphere will be maintained within the cask during the storage 

lifetime. 

5.4.5 Supplemental Information 

The reviewer should ensure all supportive information or documentation that justifies 

assumptions or analytical procedures is provided in the application. 

5.5 Review Procedures 

Figure 5-1 presents an overview of the evaluation process for coordination with other review 

disciplines. 

5.5.1 Confinement Design Characteristics (MEDIUM Priority) 

5.5.1.1 Design Criteria 

The reviewer should examine the principal design criteria presented in SAR Chapter 2 as well 

as any additional detail provided in SAR Chapter 5, “Confinement.” 
 5-5 

Figure 5-1 Overview of the Confinement Evaluation 

Chapter 1 – General 

Information Evaluation

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation

• DSS Description and 

Operational Features

• Drawings

• DSS Contents

• Accident Conditions

• Leak Tests

Chapter 12 – Accident 

Analysis Evaluation

• Event Cause

• Event Detection

• Event Consequences and 

Regulatory Compliance

• Corrective Action

Chapter 9 – Operating 

Procedures Evaluation

• Cask Loading

• Cask Handling and 

Storage Operations

Confinement Analysis

• Normal Conditions

• Off-normal Conditions

• Accident Conditions

Supplemental Information

Confinement Design Characteristics

• Design Criteria

• Design Features

Confinement Monitoring Capability Nuclides with 

Potential for Release

Chapter 5 – Confinement Evaluation

Confinement Analysis

• Normal Conditions

• Off-normal Conditions

• Accident Conditions

Supplemental Information

Confinement Design Characteristics

• Design Criteria

• Design Features

Confinement Monitoring Capability Nuclides with 

Potential for Release

Chapter 5 – Confinement Evaluation

Chapter 2 – Principal 

Design Criteria Evaluation

• General Design Criteria

• Spent Fuel Specifications

• Structural Design Criteria

• Thermal Design Criteria

• Shielding/Radiation 

Protection Design Criteria

• Operating Procedures

Chapter 4 – Thermal 

Evaluation

• Temperature

• Pressure

Chapter 6 – Shielding 

Evaluation

• Radiation Source Term

• Nuclides Available for 

Release

Chapter 8 – Materials 

Evaluation

• Closure Specifications

• Weld Design and 

Specifications

• Bolt Applications

• Seals

Chapter 3 – Structural 

Evaluation

• Spent Fuel Specifications

• General Design Criteria

• Thermal Design Criteria

Chapter 11 – Radiation 

Protection Evaluation

• Estimated Dose Rates

• Features
 5-6 

5.5.1.2 Design Features 

The reviewer should examine the general description of the cask presented in SAR Chapter 1, 

“General Description,” as well as any additional information provided in SAR Chapter 5, 

“Confinement Evaluation”. All drawings, figures, and tables describing confinement features 

should be sufficiently detailed to stand alone. 

The reviewer should verify that the applicant has clearly identified the confinement boundaries. 

This identification should include the confinement vessel, its penetrations, valves, seals, welds, 

and closure devices, and corresponding information concerning the redundant sealing. 

The reviewer should verify that the design and procedures provide for drying and evacuation of 

the cask interior as part of the loading operations. Also, the reviewer should verify that the 

confinement design is acceptable for the pressures that may be experienced during normal, offnormal and accident conditions. 

The reviewer should verify that, on completion of cask loading, the gas fill of the cask interior is 

at a pressure level that is expected to maintain a nonreactive environment and heat transfer 

capabilities for at least the 20-year storage life of the cask interior under both normal and offnormal conditions and events. This verification can include pressure testing, seal monitoring, 

and maintenance for casks with seals that are not welded if these are included in Chapter 13, 

“Technical Specifications and Operating Controls and Limits Evaluation,” of this SRP as 

conditions of use. Acceptance tests for pressure testing are described in Section 10.5.1.1, 

“Structural/Pressure Tests,” of this SRP. The NRC has previously accepted specification of an 

overpressure of approximately 14 kPa (~2 psig) and cask leak testing as conditions of use for 

satisfying this requirement. However, this general rule is not applicable to those designs that 

employ a pressurized content (i.e., to several atmospheres) to facilitate natural circulation 

cooling within the canister 

The reviewer should coordinate with the structural and materials disciplines respectively 

reviewing Chapter 3, “Structural Evaluation,” and Chapter 8, “Materials Evaluation,” of this SRP 

to ensure that the applicant has provided proper specifications for all welds and, if applicable, 

that the bolt torque for closure devices is adequate and properly specified. If applicable, the 

reviewer should verify the capability of the seal to maintain long-term closure. Because of the 

performance requirements over the 20-year license period, the reviewer should evaluate the 

potential for seal deterioration associated with bolted closures. The NRC staff has previously 

accepted only metallic seals for the primary confinement. This review should be coordinated 

with the thermal discipline to ensure that the operational temperature range for the seals 

(specified by the manufacturer) will not be exceeded. 

The staff has concluded that welded canisters can be used as a confinement system provided 

that the following design/qualification guidance is met: 

 • The canister is constructed from austenitic stainless steel. 

 • The canister closure welds meet the guidance of Section 8.5.2.3, “Weld Design 

and Specifications,” of this SRP. 

 • The canister maintains its confinement integrity during normal conditions, 

anticipated occurrences, and credible accidents and natural phenomena as 

required in 10 CFR Part 72. 
 5-7 

 • The canister shell has been helium leak tested prior its loading as required by 

10 CFR 72.236(i). This test demonstrates that the canister is free of defects that 

could lead to a leakage rate greater than the design basis leakage rate which 

could result in doses at the control area boundary in excess of the regulatory 

limits. 

 • Records documenting the fabrication and closure welding of canisters shall 

comply with the provisions of 10 CFR Part 72.174, “Quality Assurance Records” 

and SRP Section 8.5.2.3. Records storage should comply with ANSI N45.2.9, 

“Requirements for Collection, Storage, and Maintenance of Quality Assurance 

Records for Nuclear Power Plants.” 

 • Activities related to inspection, evaluation, documentation of fabrication, and 

closure welding of canisters shall be performed in accordance with a 

NRC-approved quality assurance program as required in 10 CFR Part 72, 

Subpart G, “Quality Assurance.” 

The qualification standards discussed above provide a sufficient alternative to the fabrication, 

periodic, and pre-shipment leak-testing requirements of ANSI 14.5 for the final closure welds. 

5.5.2 Confinement Monitoring Capability (LOW Priority) 

The NRC staff has found that casks closed entirely by welding do not require seal monitoring. 

However, for casks with bolted closures, the staff has found that a seal monitoring system is 

required to adequately demonstrate that seals can function to limit releases and maintain a 

helium atmosphere in the cask for the term of the 10 CFR Part 72 general license. A seal 

monitoring system, combined with periodic surveillance, enables the licensee to determine 

when to take corrective action to maintain safe storage conditions. 

Although the details of the monitoring system may vary, the general design approach has been 

to pressurize the region between the redundant seals with a nonreactive gas to a pressure 

greater than that of the cask cavity and the atmosphere. The monitoring system is leak tested 

to the same leak rate as the confinement boundary. Installed instrumentation is routinely 

checked per surveillance requirements. A decrease in pressure between these seals indicates 

that the nonreactive gas is leaking either into the cask cavity or out to the atmosphere. For 

normal operations, radioactive material should not be able to leak to the atmosphere; hence, 

this design allows for detecting a faulty seal without radiological consequence. Note that the 

volume between the redundant seals should be pressurized using a nonreactive gas, thereby 

preventing contamination of the interior cover gas. 

The staff has accepted monitoring systems as not important to safety and classified them as 

Category B under the guidelines of NUREG/CR-6407, “Classification of Transportation 

Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety 

(INEL-95/0551).” Although its function is to monitor confinement seal integrity, the failure of the 

monitoring system alone does not result in a gross release of radioactive material. It is termed 

as not important to safety since most of the associated hardware have not met the important to 

safety programmatic controls, like design, or procurement. Consequently, the monitoring 

system for bolted closures need not be designed to the same requirements as the confinement 

boundary (i.e., ASME Section III). 
 5-8 

Dependant on the monitoring system design, there could be a lag time before the monitoring 

system indicates a postulated degraded seal leakage condition. Degraded seal leakage is 

leakage greater than the tested rate that is not identified within a few monitoring system 

surveillance cycles. The occurrence of a degraded seal without detection is considered a 

“latent” condition and should be presumed to exist concurrently with other off-normal and 

design-basis events (see Section 2.5.2.2, “External Conditions,” of this SRP). Note that once 

the degraded seal condition is detected, the cask user will initiate corrective actions. 

For the “latent” condition, the monitoring system boundary would remain intact and this 

condition would be bounded by the off-normal analysis. If the monitoring system would not 

maintain integrity under design-basis accident conditions, additional safety analysis may be 

necessary. The staff recognizes that the possibility of a degraded seal condition is small and 

that the possibility of a degraded seal condition concurrent with a design-basis event that 

breaches the monitoring system pressure boundary is very remote. However, these 

probabilities have not been quantified. To address this concern, the staff accepts a 

demonstration that the dose consequences of this event are within the limits of 

10 CFR 72.106(b). 

The reviewer should examine the specified pressure of the gas in the monitored region to verify 

that it is higher than both the cask cavity and the atmosphere. The reviewer should coordinate 

with the structural and thermal reviewers associated with Chapters 3 and 4 of this SRP to verify 

the pressure in the cask cavity. 

The reviewer should examine the applicant’s analysis to verify that the total volume of gas in the 

cavity is such that normal seal leakage will not cause all of the gas to escape over the lifetime of 

the cask. The proposed maximum leakage rate should be based on the confinement evaluation 

described in Sections 5.5.3 and 5.5.4 of this SRP. The maximum allowable leakage rate should 

be specified as a minimum acceptance test criterion in SAR Chapter 9, “Acceptance Criteria 

and Maintenance Program,” and Chapter 13, “Technical Specifications and Operating Controls 

and Limits Evaluation,” even though the actual leakage rate of the seals is expected to be 

significantly lower. 

For redundant welded closures, the reviewer should ensure that the applicant has provided 

adequate justification that the welds have been sufficiently designed, fabricated, tested and 

examined to ensure that the weld will behave similarly to the adjacent parent material of the 

cask. 

The reviewer should verify that any leakage test, monitoring, or surveillance conditions are 

appropriately specified in SAR Chapter 10 “Acceptance Tests and Maintenance Program 

Evaluation”; Chapter 12, “Accident Analyses”; Chapter 13, “Technical Specifications and 

Operational Controls and Limits Evaluation” ; and/or the Certificate of Compliance (CoC). 

5.5.3 Nuclides with Potential for Release (LOW Priority) 

The quantities of radioactive nuclides are often presented in the SAR Chapter 6, “Shielding 

Evaluation,” since they are generally determined during the evaluation of gamma and neutron 

source terms in the shielding analysis. The reviewer should coordinate with the shielding 

discipline to verify that the applicant has adequately developed the source term. 

For determination of the radionuclide inventory available for release, the NRC staff has 

accepted, as a minimum for the analysis, the activity from the 60Co in the crud, the activity from 
 5-9 

iodine, fission products that contribute greater than 0.1 percent of design basis fuel activity, and 

actinide activity that contributes greater than 0.01 percent of the design basis activity. In some 

cases, the applicant may have to consider additional radioactive nuclides, depending upon the 

specific analysis. The total activity of the design basis fuel should be based on the cask design 

loading that yields the bounding radionuclide inventory (considering initial enrichment, burnup, 

and cool time). 

The staff has determined that, as a minimum, the fractions of radioactive materials available for 

release from spent nuclear fuel (SNF), provided in Table 5-2 for pressurized-water reactor 

(PWR) fuel and boiling-water reactor (BWR) fuel for normal, anticipated occurrences (offnormal) and accident-level conditions, should be used in the confinement analysis to 

demonstrate compliance with 10 CFR Part 72. These fractions account for radionuclides 

trapped in the fuel matrix and radionuclides that exist in a chemical or physical form that is not 

releasable to the environment under credible normal, off-normal, and accident-level conditions. 

Other release fractions may be used in the analysis provided the applicant properly justifies the 

basis for their usage. For example, the staff has accepted, with adequate justification, reduction 

of the mass fraction of fuel fines that can be released from the cask. Also, for the applicant to 

utilize the release fractions in Table 5-2, the reviewer should ensure that the condition of the fuel 

described in the SAR is bounded by the experimental data presented in NUREG/CR-6487. 

Specifically, this experimental data is based on the release from a single breach of one fuel rod 

and this data should not be used for spent fuel described as damaged. 

Fuel rods that are classified as damaged due to a preloading cladding breach may not have a 

driving force for the release of particulate from the rod under normal or off-normal conditions, 

providing the canister is not pressurized. However, under an impact accident damaged fuel 

rods might release additional fuel fines to the fracture of the fuel, especially the rim region in 

high burnup fuel. In addition, some canisters may be pressurized to several atmospheres and 

cask blowdown could also affect releases. Each applicant should establish release fractions for 

damaged fuel based on applicable physical data and other analyses appropriate for the specific 

type of fuel, accident impacts, and damaged condition of DSS. Alternatively, a leak-tight 

confinement boundary may be specified to preclude the release analyses of damaged fuel. 

The staff has accepted the rod breakage fractions in Section 4.5.4.6, “Pressure Analysis,” of this 

SRP for the confinement evaluations. It is important to recognize that confinement boundary 

failure under design basis normal or accident-level conditions is not acceptable. Confinement 

boundary structural integrity during design basis conditions is confirmed by the structural 

analysis. The confinement analyses demonstrate that, at the measured leakage rates and 

assumed nominal meteorological conditions, the requirements of 10 CFR 72.104(a) and 

10 CFR 72.106(b) can be met. Each independent spent fuel storage installation (ISFSI), 

whether it is a site-specific or general license, is also required to have a site-specific 

confinement analysis and dose assessment to demonstrate compliance with these regulations. 

Table 5-2 Fractions of Radioactive Materials Available for Release from Spent Fuela

Variable Fractions Available for Releaseb
 5-10 

PWR and BWR Fuel

Normal and Offnormal Conditions

Design Basis 

Accident 

Conditions

Fraction of Fuel Rods Assumed to Fail 0.01 (normal) 

0.10 (off-normal) 1 

Fraction of Gases Released Due to a 

Cladding Breach, fG

C 0.3 0.3 

Fraction of Volatiles Released Due to a 

Cladding Breach, fV

C 2 x 10-4 2 x 10-4

Mass Fraction of Fuel Released as Fines 

Due to Cladding Breach, fF

3 x 10-5 3 x 10-5

Fraction of Crud that Spalls Off Cladding, 

fC

0.15d

 1.0d

a Values in this table are taken from NUREG/CR-6487. 

b Except for Co-60, only failed fuel rods contribute significantly to the release. Total fraction of radionuclides available for 

release should be multiplied by the fraction of fuel rods assumed to have failed. 

c In accordance with NUREG/CR-6487, gases species include H-3, I-129, Kr-81, Kr-85, and Xe-127; volatile species 

include Cs-134, Cs-135, Cs-137, Ru-103, Ru-106, Sr-89, and Sr-90. 

d The source of radioactivity in crud is Co-60 on fuel rods. At the time of discharge from the reactor, the specific activity, Sc, 

is estimated to be 140 μCi/cm2

 for PWRs and 1254 μCi/cm2 for BWRs. Total Co-60 activity is this estimate times the total 

surface area of all rods in the cask (Sandoval, et al., 1991). Decay of Co-60 to determine activity at the minimum time 

before loading is acceptable.

5.5.4 Confinement Analyses (MEDIUM Priority) 

The reviewer should examine the applicant’s confinement analysis and the resulting doses for 

the normal, off-normal, and accident conditions at the controlled area boundary. 

The analysis typically includes the following common elements: 

 • Calculation of the specific activity (Ci/cm3

) for each radioactive isotope in the 

cask cavity based on rod breakage fractions, release fractions, isotopic inventory, 

and cavity free volume. 

 • Using the tested leak rate and conditions during testing as input parameters, 

calculation of the adjusted maximum seal leakage rates (cm3

/s) under normal, 

off-normal, and accident conditions (e.g., temperatures and pressures). 

 • Calculation of isotope specific leak rates (Qi-Ci/s) by multiplying the isotope 

specific activity by the maximum seal leakage rates for normal, off-normal, and 

accident conditions. 

 • Determination of doses to the whole body, thyroid, other critical organs, lens of 

the eye, and skin from inhalation and immersion exposures at the controlled area 

boundary (considering atmospheric dispersion factors -P/Q). 

The application should specify maximum allowable “as tested” seal leakage rates as a 

Technical Specification, as discussed in SRP Chapter 13. Guidance on the calculations of the 

specific activity for each isotope in the cask and the maximum allowable helium seal leakage 
 5-11 

rates for normal, off-normal, and accident-level conditions can be found in NUREG/CR-6487, 

“Containment Analysis for Type B Packages Used to Transport Various Contents” (Anderson, 

1996), and ANSI N14.5-1997. The minimum distance between the casks and the distance to 

the controlled area boundary is generally also a design criterion; however, 10 CFR 72.106(b) 

requires this distance to be at least 100m from the ISFSI. 

For the dose calculations, the NRC staff has accepted the use of either an adult breathing rate 

(BR) of 2.5x10-4 m3

/s (8.8x10-3 ft3/s), as specified in Regulatory Guide (RG) 1.109, “Calculations 

of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of 

Evaluating Compliance with 10 CFR Part 50 Appendix I,” or a worker breathing rate of 3.3x10-4

m3

/s (1.2x10-2 ft3/s), as specified in the U.S. Environmental Protection Agency (EPA) Guidance 

Report No. 11, “Limiting Values of Radionuclide Intake and Air Concentration and Dose 

Conversion Factors for Inhalation, Submersion, and Ingestion” (EPA, 1988). The dose 

conversion factors (DCF) in EPA Guidance Report No. 11 for the whole body, critical organs, 

and thyroid doses from inhalation should be used in the calculation. The bounding DCFs from 

EPA Report No. 11 should be used for each isotope unless the applicant justifies an alternate 

value. The staff is not accepting weighting or normalization of the dose conversion factors. For 

each isotope, the committed effective dose equivalent (CEDEi - for the internal whole body 

dose) or the committed dose equivalent (CDEi - for the internal organ dose) are calculated as 

follows: 

 CEDEi or CDEi (in mrem per year for normal/off-normal or mrem per accident) 

= Qi * DCFi *P/Q * BR * Duration * conversion factor (The conversion factor, if 

required, converts the input units into the desired form [CEDEi or CDEi] in mrem 

per year for normal/off-normal or mrem per accident). 

For the contributions to the whole body, thyroid, critical organs, and skin doses from immersion 

(external) exposure, the DCFs in EPA Guidance Report No. 12, “External Exposure to 

Radionuclides in Air, Water, and Soil” (EPA, 1993), should be used. Again, the NRC staff is not 

accepting weighting or normalization of the dose conversion factors. 

The deep dose equivalent (DDEi - for the external whole body) and the shallow dose equivalent 

(SDEi - for the skin dose) are calculated as follows: 

 DDEi or SDEi (in mrem per year for normal/off-normal or mrem per accident) 

= Qi * DCFi *P/Q * Duration * conversion factor2

The total effective dose equivalent, TEDE = G CEDEi + G DDEi

For a given organ, the total organ dose equivalent, TODE = G CDEi + G DDEi

The total skin dose equivalent SDE = G SDEi

Compliance with the lens dose equivalent (LDE) limit is achieved if the sum of the SDE and the 

TEDE does not exceed 0.15 Sv (15 rem). This approach is consistent with guidance in the 

Publication 26 of International Commission on Radiological Protection (ICRP), “Statement from 

the 1980 Meeting of the ICRP” (ICRP, 1980) and as specified in SRP Chapter 11, “Radiation 

Protection Evaluation.” 

In general, the staff evaluates analyses for normal, off-normal, and accident-level conditions. 

 

2

 The conversion factor, if required, converts the input units into the desired form, e.g., mrem/year. 
 5-12 

5.5.4.1 Normal Conditions 

For normal conditions, a bounding exposure duration assumes that an individual is present at 

the controlled area boundary for one full year (8,760 hours). An alternative exposure duration 

may be considered by the staff if the applicant provides justification. 

Because any potential release resulting from seal leakage would typically occur over a 

substantial period of time, the staff accepts (for applications for certificates) calculation of the 

atmospheric dispersion factors (P/Q) according to RG 1.145, “Atmospheric Dispersion Models 

for Potential Accident Consequence Assessments at Nuclear Power Plants,” assuming 

D-stability diffusion and a wind speed of 5 m/s (16 ft/s). 

For the likely case of an ISFSI with multiple casks, the doses need to be assessed for a 

hypothetical array of casks during normal conditions according to Section 2.5.3.4, 

“Shielding/Confinement/Radiation Protection,” of this SRP. Therefore, the staff anticipates that 

the resulting doses from a single cask will be a small fraction of the limits prescribed in 

10 CFR 72.104(a) to accommodate the array and the external direct dose. 

Note: If the region between redundant, confinement boundary, mechanical seals is maintained 

at a pressure greater than the cask cavity, the monitoring system boundaries are tested to a 

leakage rate equal to the confinement boundary, the pressure is routinely checked, and the 

instrumentation is verified to be operable in accordance with a Technical Specification 

Surveillance Requirement, the NRC staff has accepted that no discernible leakage is credible. 

Therefore, calculations of dose to the whole body, thyroid, and critical organs at the controlled 

area boundary from atmospheric releases during normal conditions would not be required. 

5.5.4.2 Off-Normal Conditions (Anticipated Occurrences) 

For off-normal conditions, the bounding exposure duration and atmospheric dispersion factors 

(P/Q) are the same as those discussed above for normal conditions. 

To demonstrate compliance with 10 CFR 72.104(a), the staff accepts whole body, thyroid, and 

critical organ dose calculations for releases from a single cask. However, the dose contribution 

from cask leakage should also be a fraction of the limits specified in 10 CFR 72.104(a) since the 

doses from other radiation sources are added to this contribution. 

5.5.4.3 Design-Basis Accident Conditions (Including Natural Phenomenon Events) 

For accident-level conditions, the duration of the release is assumed to be 30 days (720 hours). 

A bounding exposure duration assumes that an individual is also present at the controlled area 

boundary for 30 days. This time period is the same as that used to demonstrate compliance for 

reactor facilities licensed per 10 CFR 50 and provides good defense in depth since recovery 

actions to limit releases are not expected to exceed 30 days. 

For accident-level conditions, the staff has accepted calculation of the atmospheric dispersion 

factors (P/Q) of RG 1.145 or RG 1.25, “Assumptions Used for Evaluating the Potential 

Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage 

Facility for Boiling and Pressurized Water Reactors,” on the basis of F-stability diffusion and a 

wind speed of 1 m/s (3.3 ft/s). 
 5-13 

To demonstrate compliance with 10 CFR 72.106(b), the staff accepts whole body, thyroid, 

critical organ, and skin dose calculations for releases of radionuclides from a single cask. 

5.5.5 Supplemental Information 

The reviewer should ensure that all supportive information or documentation has been provided 

or is readily available. This includes, but is not limited to, justification of assumptions or 

analytical procedures, test results, photographs, computer program descriptions, input and 

output, and applicable pages from referenced documents. Reviewers should request any 

additional information needed to complete the review. 

5.6 Evaluation Findings 

The reviewer should examine the 10 CFR Part 72 acceptance criteria and provide a summary 

statement for each. These statements should be similar to the following model: 

 F5.1 Chapter(s) of the SAR describe(s) confinement structures, systems, and 

components (SSCs) important to safety in sufficient detail in to permit evaluation 

of their effectiveness. 

 F5.2 The design of the (cask designation) adequately protects the spent fuel cladding 

against degradation that might otherwise lead to gross ruptures. Chapter 4, 

Thermal Evaluation” of the safety evaluation report (SER) discusses the relevant 

temperature considerations. 

 F5.3 The design of the (cask designation) provides redundant sealing of the 

confinement system closure joints by . 

 F5.4 The confinement system is monitored with a monitoring system as 

discussed above (if applicable). No instrumentation is required to remain 

operational under accident conditions. 

 F5.5 The quantity of radioactive nuclides postulated to be released to the environment 

has been assessed as discussed above. In Chapter 11, “Radiation Protection 

Evaluation” of the SER, the dose from these releases will be added to the direct 

dose to show that the (cask designation) satisfies the regulatory requirements of 

10 CFR 72.104(a) and 10 CFR 72.106(b). 

 F5.6 The cask confinement system has been evaluated (by appropriate tests or by 

other means acceptable to the NRC) to demonstrate that it will reasonably 

maintain confinement of radioactive material under normal, off-normal, and 

credible accident conditions. 

A summary statement similar to the following should be made: 

“The staff concludes that the design of the confinement system of the (cask designation) 

is in compliance with 10 CFR Part 72 and that the applicable design and acceptance 

criteria have been satisfied. The evaluation of the confinement system design provides 

reasonable assurance that the (cask designation) will allow safe storage of spent fuel. 

This finding is reached on the basis of a review that considered the regulation itself, 
 5-14 

appropriate regulatory guides, applicable codes and standards, the applicant’s analysis 

and the staff’s confirmatory analysis, and accepted engineering practices.” 
 6-1 

6 SHIELDING EVALUATION 

6.1 Objective 

The shielding review evaluates the ability of the proposed shielding features to provide 

adequate protection against direct radiation from the dry storage system (DSS) contents. The 

shielding features should limit the dose to the operating staff and members of the public so that 

the dose remains within regulatory requirements during normal operating, off-normal, and 

design-basis accident (DBA) conditions. The review seeks to ensure that the shielding design is 

sufficient and reasonably capable of meeting the operational dose requirements of 10 CFR 

72.104 and 72.106 in accordance with 10 CFR 72.236(d). 

6.2 Areas of Review 

This chapter of the DSS Standard Review Plan (SRP) provides guidance for use in evaluating 

the shielding features of the proposed cask system. As defined in Section 6.5, “Review 

Procedures,” the shielding evaluation may encompass the following areas of review: 

Shielding Design Description

Design Criteria 

Design Features 

Radiation Source Definition

Gamma Source 

Neutron Source 

Shielding Model Specification

 Configuration of Shielding and Source 

 Material Properties 

Shielding Analyses

 Computer Codes 

 Flux-to-Dose-Rate Conversion 

 Dose Rates 

 Confirmatory Analysis 

Supplementary Information

 Shielding model description 

 Computer model input and output 

As prescribed in 10 CFR Part 72, the regulatory requirements for doses at and beyond the 

controlled area boundary include both direct radiation and radionuclides in effluents. An overall 

assessment of the compliance of the proposed DSS with these regulatory limits is contained in 

Chapter 11, “Radiation Protection Evaluation,” of this SRP. 

In order to ensure that the shielding design of the DSS meets the regulatory requirements as 

defined in 10 CFR Part 72, the applicant should also include information in the SAR regarding 

the technical specifications which are necessary for the DSS system to meet the dose limits at 

the controlled area boundary (See Chapter 13). 
 6-2 

In addition, the applicant should demonstrate that the system design, uses, and operating 

procedures follow the ALARA Principle. 

6.3 Regulatory Requirements 

10 CFR Part 72 requires that spent fuel storage and handling systems be designed with 

adequate shielding to provide sufficient radiation protection under normal, off-normal, and 

accident-level conditions. The DSS application should describe the design principle and 

functional features of the shielding structures, systems, and components (SSCs) important to 

safety in sufficient detail to allow the U.S. Nuclear Regulatory Commission (NRC) staff to 

thoroughly evaluate their effectiveness. It is the responsibility of the vendor and the facility 

owner to analyze such SSCs with the objective of assessing the impact of direct radiation doses 

and effluent releases to the environment on public health and safety. The reviewers should 

verify the applicant’s evaluations through review of the applicant’s model, or confirmatory 

analyses or independent modeling analysis. In addition, SSCs important to safety should be 

designed to withstand the effects of both credible accidents and severe natural phenomena 

without impairing their capability to perform their safety functions. 

This section presents a summary matrix of the portions of 10 CFR Part 72 that are relevant to 

the review areas addressed by this chapter. The NRC staff reviewer should read the exact 

referenced regulatory language. The NRC staff reviewer should verify the association of 

regulatory requirements with the areas of review presented in the matrix to ensure that no 

requirements are overlooked as a result of unique design features. Table 6-1 matches the 

regulatory requirements associated with this chapter to the areas of review. 

Table 6-1 Relationship of Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Regulations

72.104 72.106(b) 72.122(b), 

(c) 72.236(d) 

Shielding Design Description ● ●

Radiation Source Definition ● ● ● ●

Shielding Model Specification ● ● ● ●

Shielding Analyses ● ● ● ●

6.4 Acceptance Criteria 

Several technical and licensing factors should be considered during the shielding evaluation of 

the proposed DSS. First, 10 CFR Part 72 states regulatory dose limits in terms of annual sitespecific doses for normal conditions and total absorbed dose from accident conditions. 

Because the regulations do not specify cask dose rates (such as package dose rates in 10 CFR 

Part 71), site-specific factors will have to be considered at each ISFSI when determining 

compliance with the dose limits in 10 CFR 72.104 and 10 CFR 72.106. These site-specific 

factors include the geometric arrangement of storage cask arrays, topography, distances to 

dose receptors, exposure times of dose receptors, actual spent nuclear fuel (SNF) loading 
 6-3 

patterns in each storage cask, and dose contributions from other surrounding fuel cycle 

facilities. Because all of these potential site-specific factors at various sites cannot be fully 

considered in the safety analysis report (SAR) for a DSS design, the regulations in 

10 CFR 72.236(d) only require that a demonstration of the shielding design is sufficient to 

satisfy 10 CFR 72.104 and 72.106. The general licensee is required by 10 CFR 72.212 to 

consider its site-specific factors and ultimately demonstrate compliance with 10 CFR 72.104. 

Therefore, the acceptance criteria for DSS shielding seek to define standard analyses for single 

casks, and a generic array of casks, to demonstrate a sufficient shielding design. In addition, 

the acceptance criteria seek to establish acceptable dose rate levels surrounding each DSS and 

acceptable dose calculation methodologies for further use by general licensees. 

In general, the DSS shielding evaluation should provide reasonable assurance that the 

proposed design fulfills the following acceptance criteria: 

 1. The radiation shielding features of the proposed DSS are sufficient for it to meet 

the radiation dose requirements in 10 CFR 72.104 and 72.106(b). The applicant 

demonstrates this with: 

a. A shielding analysis of the surrounding dose rates that contribute to 

occupational exposure and off-site doses at large distances (for a single 

storage and transfer cask with bounding fuel source terms at various cask 

locations), and 

 b. A shielding analysis of a single cask and a generic array of casks at large 

distances. 

 2. The shielding features of and the radiations emitted by the cask, in conjunction 

with its proposed operating procedures presented in Chapter 9, “Operating 

Procedures,” of the SAR, are consistent with a well-established “as low as is 

reasonably achievable” (ALARA) program for activities in and around the storage 

site. 

 3. Radiation shielding and confinement features must be sufficient to meet the 

requirements in 10 CFR 72.106. 10 CFR 72.106(b) states: “Any individual 

located on or beyond the nearest boundary of the controlled area may not 

receive from any design basis accident the more limiting of a total effective dose 

equivalent [TEDE] of 0.05 Sv (5 rem), or the sum of the deep dose equivalent 

[DDE] and the committed dose equivalent [CDE] to any individual organ or tissue 

(other than the lens of the eye) of 0.5 Sv (50 rem). The lens dose equivalent 

[LDE] may not exceed 0.15 Sv (15 rem) and the shallow dose equivalent [SDE] 

to skin or any extremity shall not exceed 0.5 Sv (50 rem).” 

 4. The proposed shielding features should demonstrate that the DSS is capable of 

meeting the regulatory requirements prescribed in 10 CFR Part 20. 

The following sections provide additional guidance on acceptance criteria for each area of 

review for acceptability of SAR informational content and the details and method of evaluation of 

the proposed shielding features. 

6.4.1 Shielding Design Description 
 6-4 

6.4.1.1 Design Criteria 

The requirements of 10 CFR 72.104 provide dose criteria for the members of the public. 

Chapter 2, “Principal Design Criteria,” of the SAR should specify the criteria that have been 

used as a basis for protection against direct radiation. Design criteria should include the 

identification of maximum dose rates and should also be specified for occupancy areas and 

correlated with occupancy duration and distance to radiation sources. An estimate of collective 

doses (person-rem per year) should be provided for each occupancy area under various 

operations (see Chapter 11, “Radiation Protection Evaluation” of this SRP). 

The design should consider the ALARA principle. The reviewer should note that it is the 

responsibility of the general licensee using the DSS design to develop detailed procedures that 

incorporate the ALARA objectives of its site-specific radiation protection program. Further 

information on ALARA considerations is contained in the Radiation Protection Chapter. 

6.4.1.2 Design Features 

The SAR should describe the use of shielding to reduce direct radiation dose rates, and may 

consider the following: 

 • Self-shielding provided by the radioactive material being stored; 

 • Gamma and neutron shielding provided by the structural and nonstructural 

materials forming the walls and ends of the cask; 

 • Neutron capture provided by borated materials incorporated into the cask; 

 • Shielding provided by the temporary placement of water into the cask system 

during loading and unloading procedures; and 

 • Shielding provided by temporary placement of equipment and portable shields on 

and around the cask during loading and unloading procedures. 

6.4.2 Radiation Source Definition 

The SAR should describe each type of contained radiation source used as a basis for shield 

design calculations. For spent nuclear fuels, the source terms in particles/s or MeV/s should be 

described in form of either group structure or a continuous function of energy. For non-fuel 

hardware, source in Curies or Becquerel is acceptable. For contents other than fuel or non-fuel 

hardware components, isotopic composition and photon yields for each constituent should be 

specified. For confinement evaluation purposes, the physical and chemical form, source 

geometry, radionuclide content, and estimated radiation source strength should be described. 

 

The energy group structure from the source term calculation should correspond to that of the 

cross-section set of the shielding calculation. The computer methodology or database 

application used to compute source term strength should be specified. 

6.4.2.1 Gamma Sources 

The SAR should specify gamma source terms for both spent fuel and activated materials. For 

spent nuclear fuels, the source terms should be described in a format that is compatible with 
 6-5 

shielding calculation input, typically in the form of photons/s or MeV/s per energy bin. For 

assembly hardware and non-fuel hardware, source terms specified by an amount of 60Co 

activity (in Curies or Becquerel) are acceptable. For contents other than fuel or non-fuel 

hardware components, isotopic composition and photon yields for each constituent should be 

specified. A tabulated form of the radiological characteristics is acceptable. 

The SAR should include a discussion of energetic radiations created by nuclear reactions such 

as (n,γ) in the packaging materials and the contents The SAR should also provide source term 

descriptions for induced radioactivity and the bases (assumptions and analytical methods) used 

for their estimation. Alternatively, the SAR may describe the bases for excluding induced 

radioactivity source terms. 

6.4.2.2 Neutron Sources 

The SAR should describe the neutron source terms and tabulate the neutron yield by energy 

group and the bases used to determine the source terms. 

6.4.3 Shielding Model Specification 

The application should include information in the SAR relative to materials and arrangements of 

all SSCs important to safety. 

6.4.3.1 Configuration of Shielding and Source 

The SAR should describe the geometric arrangement of shielding and include illustrations that 

identify the spatial relationships among sources, shielding, and design dose rate locations. The 

SAR should clearly indicate the physical dimensions of sources and shielding materials. The 

SAR should also identify penetrations, voids, or irregular geometries that provide potential paths 

for gamma or neutron streaming. These potential streaming paths should be clearly identifiable 

on submitted drawings. The SAR should describe design features used to minimize streaming 

through these penetrations. 

The SAR should clearly state any differences between shielding features during normal or offnormal conditions and accident-level conditions. 

6.4.3.2 Material Properties 

The shielding reviewer should consult with the materials reviewer to assure that the SAR 

adequately describes the composition and geometry of the shielding materials. 

6.4.4 Shielding Analyses 

The SAR should describe the computer codes, version, computational models, data, and 

assumptions with their bases used in evaluating shielding effectiveness, and should provide 

dose rate estimates for areas of concern. The reviewer should perform confirmatory 

calculations, as necessary, to verify the results of the applicant’s shielding analyses. 

6.4.4.1 Computer Codes 

The SAR should identify the computer codes and models used in evaluating shielding for each 

significant radiation source identified in Section 6.4.2, “Radiation Source Definition,” and 
 6-6 

reference the appropriate documentation. For each computer code used, test problem solutions 

that demonstrate substantial similarity to solutions from other sources (hand calculations, 

published literature results, etc.) should be provided. A summary should be provided in the 

SAR that compares the test problem solutions in either graphical or numeric form. These 

solutions may be referenced and need not be submitted in the SAR if the references are widely 

available or have been previously submitted to the NRC for the same computer code and 

version. 

The SAR should clearly present the data used as input for computational purposes and identify 

any differences between actual material properties or physical dimensions and those used in 

the analytical method (e.g., for simplifying the computational process). The applicant should 

defend any simplifications and assumptions by showing that the approach used will result in 

conservative (bounding) estimates. 

The SAR should address calculational error in computer codes for both radiological and thermal 

source terms. Because validation data are relatively limited for burnups above 45 GWd/MTU 

(i.e., high burnup fuel), the SAR should numerically specify source term uncertainties for high 

burnup fuels. 

The SAR should determine whether source term values with uncertainties should be applied to 

the shielding, thermal, and confinement analyses, instead of nominal calculated values. In this 

determination, the SAR may consider: (1) other conservative assumptions and design margins 

in the respective analyses; (2) the maximum fuel assembly heat loads; (3) the maximum gamma 

and neutron dose rates; and (4) any measurable temperature or dose rate limitations proposed 

in the technical specifications. 

A representative computer code input file used in the shielding computation performed for the 

DSS should be included in the SAR. 

6.4.4.2 Flux-to-Dose-Rate Conversion 

The basis for the flux-to-dose-rate conversion in the shielding analysis should be stated in the 

SAR, including conversions that are done by a computer code using its own data library. The 

SAR should include a table that shows the one to one conversion factor for each energy group 

of the cask specific source term spectrum. The NRC accepts flux-to-dose-rate conversion 

factors in American National Standards Institute/American Nuclear Society Standard 6.1.1-1977 

(ANSI/ANS-6.1.1-1977). 

6.4.4.3 Dose Rates 

The SAR evaluation of shielding effectiveness should include calculated or estimated dose rates 

in representative areas around the DSS. The dose rate calculations should account for such 

factors as a minimum distance no less than100m (328 ft.), contributions from radionuclide 

releases, and other significant factors. These criteria are identified and evaluated in the 

radiation protection evaluation described in Chapter 11 of this SRP. The criteria below relate 

primarily to the completeness of information provided in the SAR. 

The SAR should clearly indicate the physical locations on and around the casks for which dose 

rate calculations have been performed. These locations should include points on or in the 

immediate vicinity of cask surfaces where workers will perform operations during loading, 

retrieval, handling, and any projected maintenance and surveillance. For storage casks with 
 6-7 

internal labyrinthine air flow passages, the SAR should include dose rate estimates for the air 

inlets and air outlets using a computer code appropriate for streaming calculations. The SAR 

should identify points that have the highest calculated dose rates. 

The SAR should include dose rate estimates for all onsite areas at which workers will be 

exposed to elevated dose rates. Dose rates within restricted areas should be calculated in 

enough detail to estimate doses received by workers performing ISFSI operations and off-site 

doses at large distances. This should be demonstrated with a standard dose-versus-distance 

curve or table for a single cask and for a generic DSS array. 

The SAR should calculate the dose rate from the cask surface for off-normal events and DBA 

conditions to ensure compliance with 10 CFR 72.104(a) and 72.106(b), respectively. The 

computational model used for these calculations should be consistent with the expected 

condition of the cask after the event. 

6.5 Review Procedures 

Figure 6-1 presents an overview of the evaluation process and can be used as a guide to assist 

in coordinating with other review disciplines. 

6.5.1 Shielding Design Description 

6.5.1.1 Design Criteria (MEDIUM Priority) 

Dose rates at the cask surface and in the vicinity of a loaded cask may vary during storage, 

transfer, and in-storage activities. While 10 CFR Part 72 establishes dose requirements for the 

ISFSI and its operation, it does not impose specific dose rate limits on the individual casks. 

Storage cask dose rates from 20 to 400 mrem/hour have been accepted in previous Part 72 

evaluations. Acceptable dose rates depend on a number of factors such as the geometry of the 

storage array, the time workers will routinely spend in the storage array for activities like 

monitoring or maintenance, the proximity to other areas frequently occupied by workers, and the 

proximity to the controlled area boundary or other public access areas. The dose requirements 

are based on 10 CFR 20.1201 for the total expected exposure to workers during anticipated 

DSS operations, and 10 CFR 72.104 for members of the public who are located beyond the 

controlled area (i.e., assumed to be at the closest boundary but, in accordance with 10CFR 

72.106(b), at least 100m from the storage cask). 
 6-8 

Figure 6-1 Overview of the Shielding Evaluation 

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation

• Dimensions

• Configuration

• Component Materials

• Content Limits

• Shielding Tests

• Shielding Material 

Specifications

Chapter 12 – Accident 

Analysis Evaluation

• Event Consequences

• DSS Design Basis

• Spent Fuel Design Basis

Chapter 3 – Structural 

Evaluation

• Deformation

• Crushing/Puncture

• Extrusion

• Displacement of Contents 

and Shielding

Chapter 11 – Radiation

Protection Evaluation

• Estimated Dose Rates

• Design Features

Shielding Analysis

• Computer Programs

• Flux-to-Dose-Rate Conversion

• Dose Rates

– Gamma

– Neutron

• Confirmatory Calculations

Shielding Model

• Configuration

• Material Properties

Radiation Source

• Gamma

• Neutron

Shielding Design

• Design Criteria

• Design Features

Chapter 6 – Shielding Evaluation

Shielding Analysis

• Computer Programs

• Flux-to-Dose-Rate Conversion

• Dose Rates

– Gamma

– Neutron

• Confirmatory Calculations

Shielding Model

• Configuration

• Material Properties

Radiation Source

• Gamma

• Neutron

Shielding Design

• Design Criteria

• Design Features

Chapter 6 – Shielding Evaluation

Chapter 4 – Thermal 

Evaluation

• Radiation Source

• Gas Inventory

Chapter 5 –

Confinement Evaluation

• Radiation Source Term

• Nuclides Available for

Release

Chapter 9 – Operating 

Procedures Evaluation

• Radiation Level Surveys

• Streaming Paths

 

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation

• Dose Rate Limits

Chapter 2 – Principal 

Design Criteria Evaluation

Chapter 1 – General 

Information Evaluation

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation

• Optional Supplement 

Shielding

• Cask Surface Dose Rate 

Limits and Corresponding 

Administrative Program
 6-9 

The shielding reviewer should coordinate with the review of SRP Chapter 2, “Principal Design 

Criteria Evaluation,” as well as review any additional shielding-related criteria. The reviewer 

should also refer to SRP Chapter 9, “Operating Procedures Evaluation,” to consider any 

expected operating procedures that would require close proximity to the cask such as cask 

equipment that should be monitored or serviced frequently. However, the evaluated dose rates 

at the side of the same cask should be reviewed to ensure that the ALARA principles are either 

engineered into the design or evoked by specific operating procedures in Chapter 9, “Operating 

Procedures Evaluation” of the SAR. 

6.5.1.2 Design Features (HIGH Priority) 

The reviewer should be familiar with the general description of the DSS presented in Chapter 1, 

“General Description,” of the SAR, as well as any additional information provided in Chapter 6, 

“Shielding Evaluation,” of the SAR. All drawings, figures, and tables describing shielding 

features should be sufficiently detailed to allow the staff to perform an in-depth evaluation. 

6.5.2 Radiation Source Definition (HIGH Priority) 

Burnup, cooling time, initial uranium loading, and initial enrichment are parameters that affect 

the total source term of SNF. The reviewer should examine the description of the design-basis 

fuel in Chapter 2, “Principal Design Criteria” of the SAR to verify that the applicant calculated the 

bounding source term. The review confirms that the applicant examined all fuel designs and 

burnup conditions for which the cask system is to be certified, to ensure that the bounding fuel 

type and values are used. Particular attention should be devoted to the combined effects of 

gamma and neutron source terms as a function of fuel burnup, cooling times, and enrichment. 

In many cases, there is no single specific enrichment-burnup combination and cooling time that 

bounds all potential cask loadings (see the analysis presented in NUREG/CR-6716). Variations 

in fuel assembly type play a secondary role for pressurized-water reactor (PWR) fuel. For 

boiling-water reactor (BWR) fuel, void fractions and channel sizes may affect the strengths of 

neutron and gamma sources. For a cask that contains spent fuel assemblies with irradiated 

burnable poison rod assemblies (BPRAs), a potential large effect is from activated component 

hardware (mainly activated cobalt in steel). Again, NUREG/CR-6716 demonstrates that for 

BPRA designs containing stainless steel, the impact on the gamma dose rate can be large. 

The design-basis radiation source term should be based on a saturation value for activation of 

cobalt impurities or on cobalt activation from a specified maximum burnup and minimum cool 

time. The reviewer should consider other activation products, as appropriate. These values 

should be bounded by those listed in the Technical Specifications. 

6.5.2.1 Initial Enrichment 

The specifications in Chapter 2, “Principal Design Criteria” of the SAR should indicate the 

maximum fuel enrichment used in the criticality analysis. For shielding evaluations, however, 

the neutron source term increases considerably with lower initial enrichment for a given burnup. 

As present in Section 3.4.1.2 of NUREG/CR-6716, as the initial enrichment decreases, the fuel 

is exposed to a larger neutron fluence to achieve the same burnup. The larger neutron fluence 

generates larger actinide content which results in larger neutron source term and secondary 

gamma source term as illustrated in NUREG/CR-6716, Section 3.4.1.2. Consequently, the SAR 

should specify the minimum initial enrichment as an operating control and limit for cask use, or 

justify the use of a neutron source term, in the shielding analysis, that specifically bounds the 

neutron sources for fuel assemblies to be placed in the cask. Because average initial 
 6-10 

enrichments typically increase with increasing burnup within the spent fuel population, the latter 

option may be used if the applicant uses low enrichments that bound the historical enrichments 

for fuels at the proposed burnups. However, the staff should not attempt to use specific source 

terms as bases for establishing operating controls and limits for cask use because these are not 

readily inspectable parameters. The fuel assembly initial enrichment, burnup, and cooling time 

are more appropriate for use as loading controls and limits. 

 

6.5.2.2 Computer Codes for Radiation Source Definition 

The reviewer should verify that the applicant determines the source terms using a computer 

code, such as ORIGEN-S (e.g., as a SAS2 sequence of Oak Ridge National Laboratory’s 

[ORNL] “SCALE” computer code package) that is well benchmarked and recognized and widely 

used by the industry. If a vendor proprietary code is used, the reviewer should check the code 

validation and verification records and procedures, preferably with sample testing problems. 

The reviewer should ensure that appropriate descriptive information, including validation and 

verification status, and reference documentation has been provided. The reviewer also should 

determine if the computer code is suitable for determining the source terms and it has been 

correctly used. Area of Applicability (AOA) is an important aspect. The reviewer should pay 

particular attention to AOA to verify if the application falls into the parameter ranges that the 

code is validated. The reviewer should determine whether the computer code is appropriately 

applied and the SAR includes verification that the chosen cross-section library is appropriate for 

the fuel specifications being considered. Many libraries are not appropriate for a burnup 

exceeding 45,000 MWd/MTU because validation data are limited at high burnups. 

The reviewer should verify that the applicant has adequately addressed calculational error and 

uncertainties of the computer codes used to determine source terms for the thermal, shielding, 

and confinement analyses. The reviewer should consider: (1) other conservative assumptions 

and design margins in the analyses; (2) the maximum fuel assembly heat loads; (3) the 

maximum gamma and neutron dose rates (including relative contributions to total); and (4) any 

measurable temperature or dose rate limits proposed for the technical specifications. 

When reviewing the source term calculations, the reviewer should also consider the factor that 

nuclide importance changes in high burnup fuels as a function of burnup and validation data. 

The data for benchmarking the calculations and computer codes is limited at high burnups. 

Additional data and information on high burnup source term issues are provided in several 

NRC-sponsored studies (DeHart, 1996; Hermann, 1998; NUREG/CR-6700, NUREG/CR-6701, 

NUREG/CR-6798.) 

6.5.2.3 Gamma Source 

The reviewer should verify that the applicant specified gamma source terms as a function of 

energy for both the spent fuel and activated hardware. If the energy group structure from the 

source term calculation differs from that of the cross-section set of the shielding calculation, the 

applicant may need to regroup the photons. Regrouping can be accomplished by using the 

nuclide activities from the source term calculation as input to a simple decay computer code 

with a variable group structure. Some applicants will convert from one structure to another 

using simple interpolation. In general, only gammas with energies from approximately 0.8 to 2.5 

MeV will contribute significantly to the dose rate through typical types of shielding; thus, 

regrouping outside this range is of a lesser importance. The reviewer should determine whether 
 6-11 

the source terms are specified per assembly, per total assemblies, or per metric ton, and ensure 

that the total source is correctly used in the shielding evaluation. 

Determining source terms for fuel assembly hardware is generally not as straightforward as for 

the SNF due to cobalt contained in the fuel assembly hardware. The potential impact on the 

gamma dose rate could be very large during the cooling times in which 60Co is the dominant 

gamma ray source (up to about 50 years) (NUREG/CR-6716). In particular, steel clad fuel 

potentially increases the cask dose rate by more than an order of magnitude over that from 

conventional Zircaloy clad fuel. The stainless steel in the BPRAs was assumed to have a 

nominal cobalt impurity level of 800 ppm, a value associated with older assembly designs. As 

presented in NUREG/CR-6716, the largest potential effect from assemblies residing in a cask 

that contains irradiated BPRAs is from activated component hardware (mainly activated cobalt 

in steel). For BPRA designs containing stainless steel, the impact on the gamma dose rate can 

be large. The effort devoted to reviewing this calculation should be based on the contribution of 

these terms to the dose rates presented in the shielding evaluation. Also, it should be noted 

whether or not the cask is intended to contain special hardware, such as control assemblies or 

shrouds, and ensured that source terms from these components are included, if applicable. The 

reviewer should confer with the Chapter 2, “Principal Design Criteria Evaluation” review team to 

make this determination. 

Depending on the cask design, neutron interactions may result in the production of high energy 

gammas near the cask surface. If this source term is not treated by the shielding analysis 

computer code, the reviewer should verify that it is determined by other appropriate means. 

As part of the source term determination, the reviewer should verify that the applicant calculates 

the quantities of certain nuclides (e.g., 85Kr, 3

H, and 129I) for use in analyzing doses from the 

release of radioactive material during postulated accidents in later sections of the SAR. These 

calculations are typically presented in Chapter 5, “Confinement,” of the SAR with the shielding 

reviewer, in coordination with the confinement reviewer, verifying the information. 

6.5.2.4 Neutron Source 

The reviewer should verify that the neutron source term is expressed as a function of energy. 

The neutron source will generally result from both spontaneous fission and alpha-n reactions in 

the fuel. Depending on the method used to determine these source terms, the applicant may 

need to independently determine in the SAR, the energy group structure. This analysis is often 

accomplished by selecting the nuclide with the largest contribution to spontaneous fission (e.g., 

244Cm) and using that spectrum for all neutrons, since the contribution from alpha-neutron 

reactions is generally small. For SNF with cooling times less than 5 years, the analysis should 

address the spectra of 242Cm and 252Cf. 

The specification of a minimum initial enrichment may be a necessary basis for defining the 

allowed contents. The reviewer should verify that the assumed minimum enrichments bounds 

all assemblies proposed for the casks in the application. Specific limits are needed for inclusion 

in the Certificate of Compliance (CoC). Lower enriched fuel, irradiated to the same burnup as 

higher enriched fuel, produces a higher neutron source. Consequently, the reviewer should 

verify that Chapter 13, “Technical Specifications and Operational Controls and Limits 

Evaluation,” of the SAR specifies the minimum initial enrichment as an operating control and 

limit for cask use. Alternately, the applicant should specifically justify the use of a neutron 

source term, in the shielding analysis, that bounds the neutron sources for fuel assemblies to be 

placed in the cask. An applicant may demonstrate that the assumed enrichment(s) bound the 
 6-12 

proposed fuel population except for possible outliers in the SNF population. This is acceptable 

if the SAR specifically requires each user to verify minimum enrichment with the Final SAR 

values, and if there are specific dose rate limits in the technical specifications. The applicant 

and the staff should not attempt to establish specific source terms as the operating controls and 

limits for cask use. 

6.5.2.5 Other Parameters Affecting the Source Term 

The reviewer should ensure the SAR contains specific information concerning reactor 

operations that affects the source term. Several NRC technical reports (specifically, 

NUREG/CR-6716, but also NUREG/CR-6700, NUREG/CR-6701, and NUREG/CR-6798) 

discuss the potential affects of other parameters not typically included as a shielding technical 

specification (e.g., moderator soluble boron concentrations, maximum poison loading, minimum 

moderator density (for BWR fuels), and maximum specific power). For example, the net impact 

of moderator density on cask dose rates is expected to be low for PWR fuels. However, the 

reviewer should be aware that the axial variation in moderator density in BWR cores can have a 

measurable effect on the axial dose rate profile of a BWR spent fuel assembly. The dose rate 

may increase near the top of the assemblies where the moderator density was the lowest. This 

is particularly important for neutron sources because reduced moderator density will harden 

neutron spectrum and hence induce more actinide production. 

6.5.3 Shielding Model Specification (HIGH Priority) 

The reviewer should verify that the applicant adequately describes the models that were used in 

the shielding evaluation for storage under normal, off-normal, and accident-level conditions. For 

example, if the cask has an external neutron shield, it should be determined whether the cask 

would be damaged by a tipover accident or degraded in a fire. Applicants should assume liquid, 

polyesters, or other resin neutron shields are not present after an accident, unless justification is 

made that they remain intact. The reviewer should confirm this analysis with the structural and 

thermal evaluation reviews of Chapter 3, “Structural Evaluation,” and Chapter 4, “Thermal 

Evaluation,” of the SAR, as appropriate. The reviewer should also confirm that the shielding 

assumptions made in dose rate calculations, for both occupational workers and the public, are 

consistent with the design criteria and design drawings. 

6.5.3.1 Configuration of the Shielding and Source 

The reviewer should examine the sketches or figures that indicate how the shielding design of 

the canister, storage overpack, and transfer cask is modeled. The reviewer should verify that 

the model dimensions and materials are consistent with those specified in the cask drawings 

presented in Chapter 1, “General Information Evaluation” of the SAR. Voids, streaming paths, 

and irregular geometries should be accounted for or otherwise treated in a conservative 

manner. In addition, the reviewer should verify that the applicant clearly states the differences, 

if any, between normal, off-normal, and accident-level conditions. 

The reviewer should verify that the applicant properly modeled the source term locations for 

both spent fuel and structural support regions (i.e., fuel assembly hardware). In some cases, 

the fuel and basket materials may be homogenized within the fuel region to facilitate the 

shielding calculations. The reviewer should watch for cases when homogenization may not be 

appropriate. For example, homogenization should not be used in neutron dose calculations 

when significant neutron multiplication can result from moderated neutrons (i.e., when 

significant amounts of moderating materials are present such as when the cask is flooded). 
 6-13 

Similarly, homogenization should not be used in configurations where significant radiation 

streaming can occur between the basket components. 

If the applicant has requested storage of damaged fuel assemblies, ensure that the applicant 

has adequately described the proposed damage assemblies. If the fuel assemblies are 

damaged to the extent that reconfiguration of the fuel into a geometry different from intact fuel 

assemblies can occur, ensure that the applicant provides appropriate close assessments for 

normal, off-normal and accident conditions. 

SNF typically has a cosine shape burnup profile along its axial length. If axial peaking appears 

to be significant, the reviewer should verify that the applicant has appropriately accounted for 

the condition. Typically, fuel gamma source terms vary proportionally with axial burnup. Fuel 

neutron source terms vary exponentially by a power of 4.0 to 4.2 (NUREG/CR-6802, 

“Recommendations for Shielding Evaluations for Transport & Storage Packages”) with axial 

burnup (NUREG/CR-6801, “Recommendations for Addressing Axial Burnup in PWR Burnup 

Credit Analyses”). In addition, the structural support regions (e.g., top and bottom end pieces 

and plenum) of the assembly should be correctly positioned relative to the SNF. These support 

regions may be individually homogenized with the basket materials when particle streaming 

through the gaps between basket components is not an issue. Generally, however, at least 

three source regions (i.e., fuel and top/bottom assembly hardware) are necessary. Some 

canisters may also employ fuel spacers to center the SNF inside the canister. 

The reviewer should verify that the SAR shows or adequately describes the locations selected 

for the various dose calculations. The reviewer should ensure that these dose points are 

representative of all locations relevant to radiation protection issues. The reviewer should pay 

particular attention to dose rates from streaming paths to which occupational workers would be 

exposed (e.g., at vent/drain port covers, lid bolts, air vents, etc.). The shielding end points 

should be noted as well (such as lead in the cask wall in relation to the assembly hardware and 

use of fuel spacers to center the fuel). See Section 6.5.4.3 for additional information regarding 

the selection of locations for dose calculations. 

6.5.3.2 Material Properties 

The reviewer should verify that the SAR provides information concerning compositions and 

densities for all materials used in the calculation model. For nonstandard materials, such as 

neutron shields, Chapter 10 of the SAR, “Acceptance Tests and Maintenance Program 

Evaluation,” should also reference the source of the data and indicate validation criteria. Many 

shielding computer codes allow the densities to be input directly in g/cm3

. If input is required in 

atoms/barn-cm the reviewer should pay particular attention to the conversion. 

The shielding reviewer should ensure that the elemental composition and density of shielding 

materials are conservatively adjusted in the shielding analyses to account for any degradation 

from aging, high temperature, accumulated radiation exposure, and manufacturing tolerances. 

The shielding reviewer should coordinate with the materials reviewer to obtain reasonable 

assurance that any degradation that may occur will not impact the safe performance of the 

shielding materials for the term proposed in the CoC application. 

6.5.4 Shielding Analyses 

6.5.4.1 Computer Codes (MEDIUM Priority) 
 6-14 

The reviewer should evaluate the computer codes or programs used for the shielding analysis. 

There are several recognized computer codes widely used for shielding analysis. These include 

computer codes that use Monte Carlo, deterministic transport, and point-kernel techniques for 

problem solution. The point-kernel technique is generally appropriate only for gammas since 

casks typically do not contain sufficient hydrogenous material to apply removal cross-sections 

for neutrons. It is also important for the reviewer to assess whether the number of dimensions 

of the computer code being applied for the shielding analysis is appropriate for the dose rates 

being calculated. Typically, NRC staff does not accept the use of one-dimensional codes for 

calculations other than shielding designs with simple cylindrical geometries. At the least, a twodimensional calculation is generally necessary. One-dimensional computer codes provide little 

information about off-axis locations and streaming paths that may be significant to determining 

occupational exposure. Even a two-dimensional calculation may not be adequate for 

determining any streaming paths if the modeled configuration is not properly established. 

These considerations in applying a particular computer code also apply to the computation of 

dose rates at the end of storage confinement casks. In some cases, the applicant will use the 

flux output from a deep-penetration shielding code as input to a large distance, skyshine code. 

The reviewer should verify that the use and interface of these codes are appropriate. 

The reviewer should be aware that the applicants often use transport or point-kernel methods to 

calculate neutron and/or gamma importance functions (unit of (mrem/hr)/(particle/s-cm2

)). 

Multiplying the importance functions by a neutron and gamma source term-per-unit length yields 

dose rates on the surface of the cask. Using the neutron and gamma importance functions, the 

applicant could determine the minimum cooling time required to meet both a decay heat limit 

and any technical specification at the maximum dose rate limit on the side of the cask. The 

reviewer, however, should pay close attention to the applicability of the importance function to 

the actual cask content, and geometry of contents and shielding. 

A valuable primer on shielding computer codes and analysis techniques has been published by 

EPRI (Broadhead, 1995). 

The computer codes given below have been previously applied for DSS source and shielding 

analysis in applications reviewed by the NRC. However, their previous use does not constitute 

generic NRC approval and, as presented above, the reviewer is cautioned that these computer 

codes can produce errors when used incorrectly. Specifically, care should be taken to ensure 

any streaming paths in the cask are appropriately determined with multi-dimensional computer 

codes under normal, off-normal, and accident-level conditions. The reviewer should also 

determine that the SAR has specified design control measures that will ensure the quality of 

computer codes used for shield analysis. 

The source of the computer codes given below vary from government sources, such as the 

Radiation Safety Information Computational Center3

 (RSICC) and other U.S. Department of 

Energy (DOE) national laboratories, to commercial shielding computer codes. It is also 

important for the reviewer to be aware that due to proliferation and security concerns, access to 

specific U.S. government-sponsored computer code packages may be restricted and special 

permission may be required when granting their use to the applicant. The applicant should use 

a computer code version that is demonstrated to be adequate for the analysis and is valid for 

the particular computational platform used to perform the analysis. Computer codes are 

periodically updated to be compatible with the latest operating system, correct errors found in 

 

3

 Radiation Safety Information Computational Center, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, 

Tennessee, 37831-6362 and on the Internet at <http://www-rsicc.ornl.gov>. 
 6-15 

prior versions, or incorporate updated methodologies. The reviewer should also consider 

whether additional confirmatory assessments and review are needed to validate the shielding 

predictions by an applicant that uses older or unsupported codes, especially in cases where 

NRC may have updated codes and no longer have the capability to directly examine 

unsupported code models from the applicant. 

The computer codes previously applied for DSS source and shielding analyses include: 

 • MicroSkyshine (air-scattering computer code); 

 • MORSE (Monte Carlo multigroup three-dimensional neutron and gamma 

transport computer code); 

 • MCBEND (Monte Carlo multigroup three-dimensional neutron and gamma 

transport computer code similar to MORSE developed by the United Kingdom 

(UK) National Radiation Protection Board (NRPB)); 

 • MCNP (Monte Carlo n-particle transport computer code maintained by Los 

Alamos National Laboratory (LANL)); 

 • RANKERN (three-dimensional point kernel gamma transport shielding computer 

code similar to QAD-CGGP); 

 • SCALE (a modular computer code system for performing standardized computer 

analyses for licensing evaluation maintained for the NRC by ORNL); 

 • SKYSHINE-II (air-scattering computer code); and 

 • STREAMING (computer code for calculation of attenuation of a gamma flux 

incident on a variety of shielding penetrations, such as ducts and voids). 

Some other shielding computer code packages available through RSICC which have potential 

application to DSS sources include: 

 • DOORS3.2 (one-, two-, and three-dimensional discrete ordinates neutron/photon 

transport code system that includes ANISN for one-dimensional, DORT for twodimensional, and TORT for three-dimensional analysis maintained by ORNL). 

 • DANTSYS (a code system maintained by the Los Alamos National Laboratory 

(LANL) that provides discrete ordinates solutions to the neutral particle transport 

equation that include ONEDANT for one-dimensional, TWODANT for twodimensional, and THREEDANT for three-dimensional multigroup discreteordinate transport analysis. 

Some of the above computer codes have been modified or improved to perform adjoint 

calculations. Examples of the computer codes with adjoint capability are as follows: 

 • DORT (part of the DOORS3.2 computer code package), 

 • A3

MCNP (Automated Adjoint Accelerated MCNP), 
 6-16 

 • MCBEND. 

The reviewer should verify that the SAR describes each of the numerical models of the 

computer codes used in the shielding evaluation. For each computer code used, the reviewer 

should ensure that an approved, validated, and verified version of the computer code is being 

applied by verifying that the following information has been provided in the SAR: 

 • The author, source, and dated version; 

 • A description of the numerical model applied in the computer code and the extent 

and limitation of its application; and 

 • Either (1) the evaluation of computer code solutions to a series of test problems, 

demonstrating substantial similarity to solutions obtained from hand calculations, 

analytical results published in the literature, acceptable experimental tests, a 

similar computer code, or benchmark problems; or (2) the specification of 

publically available references for commonly used and well-established codes 

(e.g. SCALE and MCNP) that demonstrate validation.. 

The reviewer should examine the solution comparisons provided by the SAR and determine 

whether satisfactory agreement of computer and test solutions (or resolution of deviations) is 

evident. Ideally (though not a requirement), the computer code used for evaluation of shielded 

storage containers should have been validated with actual dose rate measurements from similar 

or prototypical SNF or high-level waste storage systems. 

6.5.4.2 Flux-to-Dose-Rate Conversion (MEDIUM Priority) 

The shielding analysis computer code may perform flux-to-dose-rate conversion using its own 

data library. For the conversions, the NRC accepts the use of ANSI/ANS 6.1.1-1977. While this 

standard was revised in 1991, the NRC has not adopted the methodology given in ANSI/ANS 

6.1.1-1991 principally for two reasons. First, the 10 CFR Part 20 radiation protection 

requirements are based on fluence-to-dose conversions that are essentially the same as those 

defined by ANSI/ANS 6.1.1-1977, and are conservative relative to those of 

ANSI/ANS 6.1.1-1991. Second, neutron dose rates determined on the basis of conversions 

performed according to ANSI/ANS 6.1.1-1991 may be significantly lower than those determined 

on the basis of 10 CFR Part 20 or ANSI/ANS 6.1.1-1977. 

6.5.4.3 Dose Rates (MEDIUM Priority) 

On the basis of experience, comparison to similar systems, or scoping calculations, the reviewer 

should make an initial assessment of whether the dose rates appear reasonable and whether 

their variation with location is consistent with the geometry and shielding characteristics of the 

cask system. The following guidance pertains to the selection of points at which the dose rates 

should be calculated. 

For normal and off-normal conditions, the applicant should indicate the dose rate at all locations 

accessible to occupational personnel during cask loading, transport to the ISFSI, and 

maintenance and surveillance operations. Generally, these locations include points at or near 

various cask components and in the immediate vicinity of the cask. Example of locations 

include vent areas, trunnion areas, peak side of the cask, peak top of the cask, the canister-gap 

region, and the bottom of the transfer cask. The applicant should also calculate the dose rates 
 6-17 

at a distance of 1m from these locations because they typically contribute to occupational 

exposures. 

The application for a cask design is required by 10 CFR 72.236(d) to demonstrate that the 

shielding and confinement features of the cask are sufficient to meet the requirements in 

10 CFR 72.104 for any real individual. The real individual is an individual at or beyond the 

controlled area, For example, a real individual may be anyone living, working, or recreating 

close to the facility for a significant portion of the year. The dose to any real individual must not 

exceed the limits specified in 10 CFR 72.104 from both the storage facility and other 

surrounding fuel cycle activities. 

However, for approval of a cask design, the applicant should evaluate the shielding and 

confinement features of a single cask and a theoretical array of casks, assuming design-basis 

source terms and full-time occupancy. The applicant should also provide analyses to facilitate 

future site-specific evaluations for each general ISFSI licensee. The single cask analysis should 

identify the minimum distance that is required to meet the dose in 10 CFR 72.104. Past 

applications have shown this distance to be typically within 200m (656 ft.) of the cask. The 

applicant should include a dose or dose rate versus distance curve for a single cask to facilitate 

a site-specific evaluation for general licensees. To satisfy 10 CFR 72.106(b), dose evaluations 

should be determined at a minimum of 100m (328 ft.) distance to the closest boundary of the 

controlled area. However, the applicant may use a longer distance, provided that the longer 

distance is made a condition of use. 

The applicant should also include a dose rate-versus-distance curve for a theoretical cask array. 

The theoretical cask array should consist of at least 20 storage casks (typically in a 2x10 array), 

and may account for shadowing effect among casks. 

It is important to note that the general ISFSI licensee is permitted to use distance or additional 

engineering features, such as berms, or both, to mitigate doses to real individuals near the site. 

If such features are used in the cask SAR evaluations, they should be included in the system 

and described in the CoC. In addition, the SAR should determine the degree to which the 

normal condition dose rates could change for the identified off-normal conditions. 

As required by 10 CFR 72.212(b)(2)(i)(C), a general licensee must perform a written evaluation 

to demonstrate that the dose limits in 10 CFR 72.104 are met. An evaluation similar to that for a 

site-specific ISFSI should be performed. The licensee may use information provided in the cask 

SAR, as well as site specific information to perform the evaluation. Evaluations performed by 

the general ISFSI licensee are not reviewed for approval by NRC; however, they are subject to 

NRC inspection and must be recorded and maintained by the general licensee. 

The general licensee should establish measures in the radiological protection program, 

environmental monitoring program, and/or operating procedures to identify and reevaluate 

potential increases in exposure to the real individuals. Compliance with the dose limits in 

10 CFR 72.104 will be verified by the environmental monitoring program with direct radiation 

measurements and/or effluent measurements, as appropriate. 

The reviewer should review the technical specifications of Chapter 13 of this SRP to ensure 

appropriate requirements are addressed in the technical specifications of the cask. In addition, 

the degree to which the normal condition dose rates could change for the identified off-normal 

conditions should be verified. The need for additional calculations should be indicated in the 

Safety Evaluation Report (SER) and in the conditions set forth in the CoC. 
 6-18 

If the above dose rate criteria are satisfied, NRC accepts that the direct-dose regulatory 

requirements can also be satisfied, although the exact details needed to comply with these 

limitations will vary from ISFSI site to site. Therefore, the SAR needs to address such 

requirements only in general terms. Detailed calculations need not be presented if Chapter 13 

of the SAR, “Technical Specifications and Operational Controls and Limits Evaluation,” assigns 

ultimate compliance responsibilities to the ISFSI site licensee. 

In addition, the applicant should calculate the dose rate at 100m (328 ft.) from the cask surface 

for accident-level conditions to assist in demonstrating the design is sufficient to meet the 

requirements of 10 CFR 72.106. The model used for these calculations should be consistent 

with the expected condition of the cask after an accident or natural event. 

The potential reconfiguration of damaged fuel within the damaged-fuel can, if applicable, must 

be analyzed to demonstrate that the cask/fuel meet the dose limits of normal and design basis 

events of storage. The shielding analysis should assume a worst case or bounding 

configuration of the canned fuel. 

6.5.4.4 Confirmatory Calculations (HIGH Priority) 

The reviewer should independently evaluate the dose rates in the vicinity of the cask for normal, 

off-normal, and accident-level conditions. In determining the level of effort appropriate for these 

calculations, the reviewer should consider the following factors: 

 • the degree of sophistication in the SAR analysis; 

 • a comparison of SAR dose rates with those of similar casks that have previously 

been reviewed, if applicable; 

 • the typical variation in dose rates expected between different computer codes 

and cross-section sets; 

 • the fact that actual dose rates will be monitored and limited by the requirements 

of 10 CFR Part 20; 

• the restrictions to be placed on the DSS operations or the limits to be placed on 

dose rates, as documented in the CoC and/or technical specifications.

 • the applicant’s experience in using the methods and computer codes in previous 

submittals; 

 • the use of new, or previously reviewed, computational methods or computer 

codes; and, 

 • the inclusion in the design of any significant departures from previous cask 

system designs (e.g., unusual shield geometry, new types of materials, or 

different source terms). 

At a minimum, the review should include examination of the applicant’s input to the computer 

code used for the shielding analysis. The reviewer should verify use of proper dimensions, 
 6-19 

material properties, and an appropriate cross-section set. In addition, the reviewer should 

independently evaluate the use of gamma and neutron source terms. 

If a more detailed review is required (e.g., a new and not previously reviewed shielding 

computer code), the reviewer should independently confirm the dose rates to ensure that the 

SAR results are reasonable and conservative. As previously noted, the use of a simple 

computer code for neutron calculations often does not provide results with sufficient accuracy 

and confidence. An extensive and more detailed evaluation may be necessary if large 

uncertainties are suspected. To the degree possible, the use of a different shielding computer 

code with a different analytical technique and cross-section set from that used in the SAR 

analysis will usually provide a more independent evaluation. 

A good reference regarding the treatment of uncertainty in thick-shielded cask analyses is the 

Electric Power Research Institute’s “Evaluation of Shielding Analysis Methods in Spent Fuel 

Cask Environments,” published in 1995 (Broadhead, 1995). 

6.5.5 Supplemental Information 

Supplemental information can include copies of applicable references (especially if a reference 

is not generally available to the reviewer), computer code descriptions, input and output files, 

and any other information that the applicant deems necessary. Likewise, the reviewer should 

request any additional information needed to complete the review process. 

6.6 Evaluation Findings 

The reviewer should review the 10 CFR Part 72 acceptance criteria and provide a summary 

statement for each. These statements should be similar to the following model: 

 F6.1 Section(s) of the SAR describe(s) shielding structures, systems, and 

components (SSCs) important to safety in sufficient detail to allow evaluation of 

their effectiveness. The reviewer should cite specific drawings that are used to 

define the SSCs for shielding. 

 F6.2 Section(s) of the SAR provide reasonable assurance that the radiation 

shielding features are sufficient to meet the radiation protection requirements of 

10 CFR Part 20, 10 CFR 72.104 and 10 CFR 72.106. 

 F6.3 Operational restrictions to meet dose and ALARA requirements in 10 CFR 

Part 20, 10 CFR 72.104, and 10 CFR 72.106 are the responsibility of the site 

licensee. The [cask designation] shielding features are designed to assist in 

meeting these requirements. 

A summary statement similar to the following should be made: 

“Based upon its review, the staff has reasonable assurance that the design of the 

shielding system of the [cask designation] is in compliance with 10 CFR Part 72 and that 

the applicable design and acceptance criteria have been satisfied. The evaluation of the 

shielding system design provides reasonable assurance that the [cask designation] will 

allow safe storage of spent fuel in accordance with 10 CFR 72.236(d). This finding is 

reached on the basis of a review that considered the regulation itself, appropriate 
 6-20 

regulatory guides, applicable codes and standards, accepted engineering practices, and 

the statements and representations in the application. 
 7-1 

7 CRITICALITY EVALUATION 

7.1 Review Objective 

The criticality review and evaluation ensures that spent nuclear fuel (SNF) to be placed into the 

dry storage system (DSS) remains subcritical under normal, off-normal, and accident conditions 

involving handling, packaging, transfer, and storage. The criticality review is designed to fulfill 

the strategic outcome of no inadvertent criticality events, part of the strategic goal of safety 

described in the agency's strategic plan (NUREG-1614). 

7.2 Areas of Review 

This portion of the DSS review evaluates the criticality design and analysis related to SNF 

handling, packaging, transfer, and storage procedures for normal, off-normal, and accident 

conditions. Consequently, this chapter of the DSS Standard Review Plan (SRP) provides 

guidance for use in conducting a comprehensive criticality evaluation that may encompass any 

or all of the following areas of review: 

Criticality Design Criteria and Features

Fuel Specification

 Non-Fuel Hardware 

 Fuel Condition 

Model Specification

 Configuration 

 Material Properties 

Criticality Analysis

 Computer Codes 

 Multiplication Factor 

 Benchmark Comparisons 

Burnup Credit

 Limits for the Licensing Basis 

 Code Validation 

 Licensing-Basis Model Assumptions 

 Loading Curve 

 Assigned Burnup Loading Value 

 Estimate of Additional Reactivity Margin 

Supplemental Information

7.3 Regulatory Requirements 

SNF storage systems must be designed to remain subcritical unless at least two unlikely 

independent events occur. Moreover, the SNF cask must be designed to remain subcritical 

under all credible conditions. Regulations specific to nuclear criticality safety of the cask system 

are specified below. Normal and accident conditions to be considered are also identified in U.S. 

Code of Federal Regulations (CFR) Part 72, “Licensing Requirements for the Independent 

Storage of Spent Nuclear Fuel and High-Level Radioactive Waste,” Title 10, “Energy” (10 CFR 
 7-2 

Part 72). The reviewer should read the exact regulatory language. Table 7-1 matches the 

relevant regulatory requirements associated with this chapter to the areas of review. 

Table 7-1 Relationship of Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Regulations

72.124 72.236(a) 72.236(b), (c), 

(g), (h), (m), 

Criticality Design Criteria and Features ● ● ●

Fuel Specification ● ● 

Model Specification ● ● ●

Criticality Analysis ● ● ●

Burnup Credit ● ● 

7.4 Acceptance Criteria 

In general, the DSS criticality evaluation seeks to ensure that a subcritical condition is 

maintained for the given design by fulfilling the following acceptance criteria: 

 • The effective neutron multiplication factor, keff, including all biases and 

uncertainties at a 95-percent confidence level, should not exceed 0.95 under all 

credible normal, off-normal, and accident-level conditions. 

 • At least two unlikely, independent, and concurrent or sequential changes to the 

conditions essential to criticality safety, under normal, off-normal, and accidentlevel conditions would need to occur before an accidental criticality is deemed to 

be possible (i.e., double contingency principle). 

 • When practicable, criticality safety of the design should be established on the 

basis of favorable geometry, permanently fixed neutron-absorbing materials 

(poisons), or both. Where solid neutron-absorbing materials are used, the design 

should provide for a positive means to verify their continued efficacy during the 

storage period. The neutron-absorbing materials’ continued efficacy may be 

confirmed by a demonstration or analysis before use, showing that significant 

degradation of these materials cannot occur over the life of the facility. 

 • Criticality safety of the cask system should not rely on credit for more than 75 

percent of the neutron poison material in fixed neutron absorbers when subject to 

standard acceptance tests. For greater credit allowance, special, comprehensive 

fabrication tests capable of verifying the presence and uniformity of the neutron 

absorber are needed. 
 7-3 

7.5 Review Procedures 

The interrelationship of the criticality evaluation review with other disciplines is shown in Figure 

7-1. The figure shows that this review draws upon information from the general information 

section as well as information reviewed or developed for the design criteria, structural, and 

operating procedures evaluations. Information collected or developed during the review of this 

chapter is useful in the evaluation of the materials, operating procedures, acceptance tests and 

maintenance program, accident analysis, and technical specifications and operating controls for 

the DSS. 

The reviewer should examine the criticality design features and criteria in SAR Chapter 1, 

“General Information,” and SAR Chapter 2, “Principal Design Criteria,” in addition to SAR 

Chapter 7, “Criticality Evaluation,” for any additional details concerning criticality design features 

and criteria. The reviewer should assess the bounding specifications for the SNF and assure 

consistency with the models used by the applicant in the criticality analyses. The reviewer 

should verify that criticality safety considerations under normal, off-normal, and accident-level 

conditions are addressed by the applicant and that the cask system design complies with 

10 CFR Part 72. In addition, the reviewer should verify that the criticality calculations determine 

the highest keff that might occur for all loading states under normal, off-normal, and accident 

conditions involving handling, packaging, transfer, and storage. To the extent practicable, the 

use of independent methods to perform any keff calculations by the reviewer should be pursued 

to evaluate the applicant’s design. 

7.5.1 Criticality Design Criteria and Features (HIGH Priority) 

The reviewer should examine the principal criticality design criteria presented in SAR Chapter 2 

as well as any related details provided in SAR Chapter 7, “Criticality Evaluation”. The general 

cask description presented in SAR Chapter 1 should be examined for any relevant information. 

The information in Chapter 7 of the SAR should be verified to be consistent with the information 

in SAR Chapters 1 and 2. The reviewer should verify that all descriptions, drawings, figures, 

and tables are sufficiently detailed to support an in-depth staff evaluation. 

The criticality design of the cask relies on the general dimensions of the cask components and 

the spacing of the fuel assemblies. The criticality design also often relies on neutron poisons. 

These may be in the form of fixed poisons in the basket structure, which may be used together 

with flux traps, and/or soluble poisons in the water of the SNF pool. During loading and 

unloading operations, NRC staff accepts the use of borated water as a means of criticality 

control if the applicant specifies a minimum boron content and strict controls are established to 

ensure that the minimum required boron concentration is maintained. This condition in turn 

becomes an operating control and limit in SAR Chapter 13, and in the Technical Specification 

(TS). The SER should also discuss these operating controls. Other design features significant 

to the criticality design, such as important basket dimensions that control the spacing of the fuel 

assemblies should also be included in the TS. These dimensions may be a minimum pitch for 

the basket cells or a minimum flux trap width. 

If borated water is used for criticality control during loading and unloading operations, 

administrative controls and/or design features should be implemented to ensure that accidental 

flooding with unborated water cannot occur, or the criticality evaluation should consider 

accidental flooding with unborated water. If the cask is also intended for transport, borated 

water should not be relied upon for criticality control. Borated water and any other liquids are 

not acceptable as a means of criticality control for a cask in dry storage. 
 7-4 

Figure 7-1 Overview of Criticality Evaluation 

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation

• DSS Contents

• Acceptance Tests

– Neutron Absorber 

Tests

Chapter 12 – Accident 

Analysis Evaluation

• Cause of Event

• Detection of Event

• Summary of Event 

Consequences

• Design Basis for 

Structures, Systems, and 

Components Important to 

Safety

– Spent Fuel 

Specifications

• Design Criteria Safety 

Protection Systems

– Criticality

Chapter 3 – Structural 

Evaluation

• Design Criteria and 

Features

Criticality Analysis

• Computer Codes

• Multiplication Factor

• Benchmark Comparisons

Burnup Credit

Burnable Neutron Absorber Credit

Supplemental Information

Criticality Design Criteria and Features

Fuel Specification

Model Specification

• Configuration

• Material Properties

Chapter 7 – Criticality Evaluation

Chapter 8 – Materials 

Evaluation

• Fuel Specifications

• Material Properties

Chapter 9 – Operating 

Procedures Evaluation

• Loading Operations

– Misloading

• Loading Operations

– Fuel Specifications

– Damaged Fuel

– Draining and Drying

• Unloading Operations

– Damaged Fuel

– Cooling, Venting, and 

Reflooding

• Subcriticality Features

Chapter 2 – Principal 

Design Criteria Evaluation

Chapter 1 – General 

Information Evaluation

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation

• Limiting Conditions

• Design Features

• Fuel (or Contents) 

Specifications

Chapter 9 – Operating 

Procedures Evaluation

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation

• DSS Contents

• Acceptance Tests

– Neutron Absorber 

Tests

Chapter 12 – Accident 

Analysis Evaluation

• Cause of Event

• Detection of Event

• Summary of Event 

Consequences

• Design Basis for 

Structures, Systems, and 

Components Important to 

Safety

– Spent Fuel 

Specifications

• Design Criteria Safety 

Protection Systems

– Criticality

Chapter 3 – Structural 

Evaluation

• Design Criteria and 

Features

Criticality Analysis

• Computer Codes

• Multiplication Factor

• Benchmark Comparisons

Burnup Credit

Burnable Neutron Absorber Credit

Supplemental Information

Criticality Design Criteria and Features

Fuel Specification

Model Specification

• Configuration

• Material Properties

Chapter 7 – Criticality Evaluation

Chapter 8 – Materials 

Evaluation

• Fuel Specifications

• Material Properties

Chapter 9 – Operating 

Procedures Evaluation

• Loading Operations

– Misloading

• Loading Operations

– Fuel Specifications

– Damaged Fuel

– Draining and Drying

• Unloading Operations

– Damaged Fuel

– Cooling, Venting, and 

Reflooding

• Subcriticality Features

Chapter 2 – Principal 

Design Criteria Evaluation

Chapter 1 – General 

Information Evaluation

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation

• Limiting Conditions

• Design Features

• Fuel (or Contents) 

Specifications

Chapter 9 – Operating 

Procedures Evaluation
 7-5 

This includes use of any credit in the criticality analysis for the presence of a liquid that may 

provide neutron shielding (and is external to the fuel basket); however, its presence and most 

reactive density should be assumed if it increases keff. Also, if more than one certified or 

licensed basket design of the same supplier could fit in the cask; the type of basket to be used 

with the cask should be stamped in a location on the cask system in a way that allows for easy 

identification of the basket. Thus, a licensee using the cask system will be able to easily verify 

the appropriateness of the fuel contents to be loaded in the basket. 

7.5.2 Fuel Specification (HIGH Priority) 

The reviewer should examine the specifications for the ranges or types of SNF that will be 

stored in the cask as presented in SAR Chapters 1, “General Information Evaluation” and 2, 

“Principal Design Criteria Evaluation” as well as any related information provided in SAR 

Chapter 7,”Criticality Evaluation”. The SNF specifications given in Chapter 7 of the SAR should 

be consistent with, or bound, the specifications given in SAR Chapters 1 and 2 and in the TS. 

The reviewer should also, keeping in mind that some specifications are more important than 

others, identify the specifications that are keys to criticality safety and verify that these are 

appropriately captured in the TS. NUREG-1745 provides a listing of some fuel specifications 

that may be keys to maintaining the system subcritical. 

Of primary interest is the type of fuel assemblies and maximum fuel enrichment that should be 

specified and used in the criticality calculations. Some boiling-water reactors (BWR) use 

multiple fuel pin enrichments, in which case the criticality calculations should use the maximum 

fuel pin enrichment present. Depending upon the fuel design, an applicant may propose use of 

assembly averaged or lattice averaged enrichments. This may be acceptable if the applicant 

can demonstrate that the applicant’s averaging technique is technically defensible and, for the 

criticality calculation, produces realistic or conservative results. Because of the natural uranium 

blankets present in many BWR designs, use of an assembly-averaged enrichment that includes 

the blankets is not normally considered appropriate or conservative for BWR fuel. 

Another parameter of interest is the fuel density assumed in the analysis. The value of the fuel 

density used in the calculations should be justified to be realistic or conservative. 

Although the burnup of the fuel affects its reactivity, many criticality analyses have assumed the 

cask to be loaded with fresh fuel (the fresh fuel assumption). Alternatively, the NRC staff has 

provided guidance for limited burnup credit for intact fuel. This guidance is currently limited to 

burnup credit available from actinide compositions associated with UO2 fuel of 5.0 wt percent or 

less enrichment that has been irradiated in a PWR to an assembly-average burnup value not 

exceeding 50 GWD/MTU and cooled out-of-reactor for a time period between 1 and 40 years. 

Guidance regarding the review of a criticality analysis that involves burnup credit is provided in 

Section 7.5.5. Specifications for the fuel that will be stored in the cask, including those 

important for burnup credit, if applicable, should be included in Chapter 13, “Technical 

Specifications and Operational Controls and Limits Evaluation” of both the SAR and SER, with 

those specifications determined to be key to criticality safety also explicitly listed in the 

Technical Specifications. 

For analyses that use the fresh fuel assumption, inadvertent loading of the cask with 

unirradiated fuel is not a major concern. However, inadvertent loading of the cask with 

unirradiated fuel is a major concern for casks that rely on criticality analyses that use burnup 

credit. Therefore, detailed loading procedures for these casks will need to include steps to 

prevent misloading of unirradiated fuel. Regardless of which analysis is used, detailed loading 
 7-6 

procedures may need to include steps to prevent misloading if fuel exceeding the design basis 

for the DSS is present in the pool at the time of loading. 

Because casks are typically designed to store many types and configurations of fuel 

assemblies, the applicant should demonstrate that criticality requirements are satisfied for the 

most reactive case. A determination of which fuel is bounding in a criticality analysis depends 

on many factors and usually requires examination of several types of fuel assemblies and 

compositions. The design-basis fuel has often been the Westinghouse 17x17 optimized fuel 

assembly (OFA); however, this will not be the case for all cask designs because of cask-specific 

effects on reactivity. Therefore, the applicant should demonstrate and reviewers should verify 

that the fuel assembly used as the design basis is the most reactive for the specific cask design. 

Chapter 1, “General Information Evaluation” of the SAR and Chapter 13, “Technical 

Specifications and Operation Controls and Limits Evaluation” of the SER should either clearly 

indicate the design-basis assemblies or reference the SAR chapter in which they are identified. 

7.5.2.1 Non-Fuel Hardware 

Some fuel assemblies may also have non-fuel components that are positioned or operated 

within the envelope of the fuel assembly during reactor operation that an applicant may seek to 

store with the assemblies in the cask. These items include PWR control assemblies such as 

Rod Cluster Control Assemblies (RCCAs), Control Element Assemblies (CEAs), Burnable 

Poison Rod Assemblies (BPRAs) and Axial Power Shaping Rods (APSRs). Applicants may 

also seek approval of storage of fuel assemblies with other items that extend into an assembly’s 

active fuel region, such as stainless steel rod inserts used to displace water in PWR assembly 

guide tube dashpots. For applications that propose to load assemblies containing non-fuel 

hardware, ensure that the analysis considers the effects of both inclusion and neglect of nonfuel hardware on system reactivity. If the application relies on the presence of the non-fuel 

hardware to meet the subcritical criterion, verify that the non-fuel hardware will remain in place 

under all normal and design basis conditions. 

Generally, staff does not allow reliance on, or credit for, fuel-related burnable neutron 

absorbers. This restriction includes residual neutron-absorbing material remaining in the nonfuel hardware loaded with an assembly. However, credit for any negative reactivity for this latter 

absorbing material may be accepted if: (1) the remaining absorbing material content is 

established through physical measurement, where a sufficient margin of safety is included 

commensurate with the uncertainty in the method of measurement, (2) the axial distribution of 

the poison depletion is adequately determined with appropriate margin for uncertainties, and 

(3) adequate structural integrity and placement of the non-fuel hardware under accident 

conditions is demonstrated. Ensure that the fuel specifications, described in Chapter 13, 

“Technical Specifications and Operation Controls and Limits Evaluation” of both the SAR and 

SER, include the important details about the non-fuel hardware to be stored with the fuel 

assemblies and the associated residual neutron absorbing material, with those details key to 

criticality safety included in the TS, as appropriate. Also, verify that operating procedures are 

established that ensure that non-fuel hardware loaded with assemblies meets the approved 

specifications as well as remains in position. 

7.5.2.2 Fuel Condition 

Determine if the applicant has included any specifications regarding the fuel condition. To date, 

a number of applications have requested approval for storage of fuel that is damaged as well as 

intact, or undamaged. The reviewer should consult the most current staff guidance for detailed 
 7-7 

descriptions regarding what constitutes damaged, undamaged and intact fuel (e.g., 

Sections 8.4.17.2 and 8.6 of this SRP or more recent guidance). This guidance gives the 

applicant the latitude to define fuel with defects (such as missing rods but not loose rods or 

debris) as undamaged fuel as long as the fuel can meet all the fuel specific or system related 

functions. For purposes of the criticality function, undamaged fuel is fuel that: (1) is in the form 

of an assembly, (2) has structural and material properties such that the assembly can withstand 

normal and design basis events while maintaining its geometric configuration and (3) has had 

any damaged or missing fuel rods replaced with solid dummy rods that displace an equal 

amount of water as the original rods. Fuel that cannot meet these criteria is considered to be 

damaged. However, a fuel assembly with missing fuel rods may be considered undamaged fuel 

if analyses are performed that show the criterion for subcriticality will be met with the fuel rods 

missing. 

A fuel assembly that is classified as damaged must be placed in a damaged fuel canister, or in 

an acceptable alternative, for loading into the cask. For a cask that is also intended for 

transport, it must be kept in mind that the more severe conditions of transport may require 

re-analysis of assemblies classified as undamaged under storage-only conditions prior to 

transport. Specifications concerning the condition of the fuel to be stored in the cask and the 

loading of damaged fuel, as applicable, should be included in Chapter 13, “Technical 

Specifications and Operation Controls and Limits Evaluation” of both the SAR and SER and in 

the Certificate of Compliance (in the TS). 

The reviewer should verify that the criticality analysis addresses the conditions of the fuel to be 

stored in the cask system. Analyses for cask systems designed to store damaged fuel should 

bound the configuration of the damaged fuel assemblies under all credible normal and design 

basis conditions. For example, some analyses have performed calculations that model the 

damaged fuel as arrays of bare fuel rods (i.e., the cladding is assumed to be completely 

removed) having an optimized rod pitch. 

7.5.3 Model Specification (HIGH Priority) 

Manufacturing and fabrication tolerances should be specified, and the reviewer should verify 

that the applicant used the most reactive combination of tolerances, within the ranges of their 

acceptable values, in the cask system model. 

7.5.3.1 Configuration 

The reviewer should verify that the model used in the criticality evaluation is adequately 

described for normal, off-normal, and accident conditions. The reviewer should also coordinate 

with the structural, materials, and thermal reviewers to understand any damage that could result 

from accident or natural phenomena events. 

The reviewer should examine the sketches or figures of the model used for criticality 

calculations. The reviewer should verify that the dimensions and materials of the model are 

consistent with the engineering drawings. Differences between the actual cask configuration 

and the models should be identified, and the models should be shown to be conservative. 

Substitution of end sections and support structures of the fuel with ordinary water is a common 

and usually conservative practice in criticality analysis. However, substitution with borated 

water is typically not conservative. Any such substitutions should be justified. 
 7-8 

Tolerances for poison material dimensions and/or concentrations should be defined, and the 

most reactive conditions should be used in the criticality analysis. In addition, the analysis 

should identify all important design conditions and then address these conditions for potential 

variations during normal, off-normal, and accident-level conditions. 

The reviewer should verify that the applicant has considered deviations from nominal design 

configurations. The evaluation of keff should not be limited to a model in which all of the fuel 

bundles are neatly centered in each basket compartment with the center line of the basket 

coincident with the center line of the cask. For example, a cask with steel confinement and lead 

shielding may have a higher keff when the basket and fuel assemblies are positioned as close as 

possible to the lead. However, in some designs, the most reactive configuration may be when 

all fuel assemblies are shifted toward the center of the basket. 

In addition to a fully flooded cask, the SAR should address configurations in which the cask is 

filled with partial density water or is partially filled with water (borated, if applicable) and the 

remainder of the cask is filled with steam consisting of ordinary water at partial density. These 

configurations are considered to be possible during loading and unloading operations. The SAR 

should also consider the possibility of preferential or uneven flooding within the cask, if such a 

scenario is credible for the given cask design (e.g., because of blockage in small flow or drain 

paths). In particular, the reviewer should watch for situations where there is water in the fuel 

regions but not in the flux traps, if applicable. Cask designs for which this type of flooding is 

credible are generally unacceptable. The SAR should also consider flooding in the fuel rod 

pellet-to-clad gap regions with unborated water. Above all, the analysis must demonstrate that 

the cask remains subcritical for all credible conditions of moderation. 

The reviewer should examine whether the applicant has prepared a heterogeneous model of 

each fuel rod or has homogenized the entire fuel assembly. With current computational 

capabilities, homogenization is now an uncommon practice and should not be used. 

7.5.3.2 Material Properties 

The reviewer should verify that the compositions and densities are provided for all materials 

used in the calculational model. The applicant should also cite, in the SAR Chapter 8, 

“Materials Evaluation”, the source of all materials data, particularly the data for fuel and poison 

materials. In coordination with the materials reviewer, the criticality reviewer should determine 

the acceptability of the sources of data that are important to the criticality safety function of the 

cask. The criticality reviewer should, in coordination with the materials reviewer, ensure that the 

applicant addressed the validation of the poison concentration in the acceptance testing 

discussion in SAR Chapter 10, “Acceptance Tests and Maintenance Program Evaluation.” 

Criticality computer codes generally will allow the densities to be input directly in units of g/cm3

or units of atoms/barn-cm. In either case, the reviewer should pay attention to the final value 

used directly by the code. Also, the reviewer should confirm that the analysis does not take 

credit for more than the minimum amount of neutron absorber verified by the acceptance 

testing, subject to the criteria in Section 7.4. 

 Among other specifications, 10 CFR Part 72 requires that a positive means to verify the 

continued efficacy of solid neutron-absorbing materials should be provided when these 

materials are used. The criticality reviewer should verify that the neutron flux from the irradiated 

fuel results in a negligible depletion of poison material over the storage period, In coordination 

with the materials and structural reviewers, the criticality reviewer should ensure that the 

applicant demonstrates that the required acceptance testing of the poisons during fabrication 
 7-9 

(specified in SAR Chapter10, “Acceptance Tests and Maintenance Program Evaluation”) has 

been satisfactorily specified, and by analysis or demonstration, the applicant has shown the 

poison material’s durability and resistance to degradation during the storage period. 

 

The neutron flux used for this analysis should be the maximum that may be produced by 

feasible loadings of irradiated or unirradiated fuel. The reviewer should coordinate review of the 

applicant’s acceptance testing and assessment of the poison material’s durability with the 

materials reviewer to verify that the applicant provides a valid and accurate demonstration of the 

absorber material’s continued efficacy. Consideration should be given to the effects of physical 

and chemical actions as well as irradiation (gamma and neutron). There may be other ways to 

provide positive means of verifying the neutron absorber’s continued efficacy. For applications 

that propose an alternative method, the reviewer should verify that the proposed method is 

reasonable (considering any effects on meeting confinement, shielding, or other system design 

criteria) and valid and accurate in demonstrating the absorber’s continued efficacy. 

7.5.4 Criticality Analysis (Priority as indicated) 

7.5.4.1 Computer Codes 

(MEDIUM Priority) Both Monte Carlo and deterministic computer codes may be used for 

criticality calculations. Monte Carlo computer codes are better suited to three-dimensional 

geometry and, therefore, are more widely used to evaluate spent fuel cask designs. The most 

frequently used Monte Carlo codes are SCALE/KENO (ORNL, 2005), MCNP (MCNP5, 2003), 

and MONK (AEA Technology, 2001). All three codes permit the use of either multigroup or 

continuous cross sections. The reviewer should determine that the applicant has used a 

computer code that is appropriate for the particular application and has used that code correctly. 

(LOW Priority) The reviewer should determine whether the applicant has chosen an acceptable 

set of cross sections. Cross sections may be distributed with the criticality computer codes or 

developed independently from another source. The applicant should provide or reference the 

source of cross-section data. For user-generated cross sections, the applicant should specify 

the method used to obtain the actual data employed in the criticality analysis. For multigroup 

calculations, the neutron flux spectrum used to construct the group cross sections should be 

similar to that of the cask. If a multigroup treatment is used, the reviewer should ensure the 

applicant has appropriately considered the neutron spectrum of the cask. In addition to 

selecting a cross-section set collapsed with an appropriate flux spectrum, a more detailed 

processing of the energy-group cross sections is required to properly account for resonance 

absorption and self-shielding. The use of multigroup KENO as part of the CSAS sequences in 

SCALE will directly enable appropriate cross-section processing. Some cross-section sets 

include data for fissile and fertile nuclides (based on a potential scattering cross section, sp) that 

can be input by the user. If the applicant has used a stand-alone version of KENO, the reviewer 

should ensure that potential scattering has been properly considered. Furthermore, information 

has been published concerning problems with some cross-section libraries once commonly 

distributed with SCALE/KENO. One library, the “working-format” library, was used for 

calculations of the code manual’s sample problems but is not intended for criticality calculations 

of actual systems (IN 91-26, 1991). Another library, the SCALE 123-group library, has 

demonstrated inadequacies for non-thermalized, highly enriched systems (NUREG/CR-6328, 

1995). 
 7-10 

MEDIUM Priority) The reviewer should pay particular attention to the proper selection of 

scattering cross section data for important compounds that may be in the system. Use of a free 

atom cross section for nuclides in a compound may not adequately account for the scattering 

effects of atoms bound in molecules and lattices. This misrepresentation can cause the 

underprediction of keff, particularly in the case of a well moderated system where energetic up 

scattering plays a significant role in the neutronics of the system. 

(MEDIUM Priority) For analyses of a cask model with separate regions of water and steam, the 

use of a multigroup cross-section set raises additional concerns. The reviewer should verify 

that the applicant has addressed the differences in the flux spectra in the two regions. If the 

results of these calculations indicate that keff is close to 0.95, additional independent calculations 

using a different code and/or cross-section library (a library derived from a different crosssection database if possible and appropriate) may be helpful. The reviewer should also closely 

examine the applicant’s benchmark analysis to verify the applicability of the critical experiments 

considered. 

7.5.4.2 Multiplication Factor 

(MEDIUM Priority) The reviewer should examine the results and discussion of the keff

calculations for the storage cask. The reviewer should verify that the calculations determine the 

highest keff that might occur during all operational states under normal, off-normal and accident 

conditions. Sensitivity parametric analyses may be used to provide the required demonstration 

that the highest keff with a confidence level of 95 percent has been determined. Variations in the 

results caused by differences in the models and sensitivity analyses should be explained and 

found to be reasonable. 

(MEDIUM Priority) For Monte Carlo calculations, the reviewer should assess if the number of 

neutron histories and convergence criteria are appropriate. As the number of neutron histories 

increases, the mean value for keff should approach a fixed value, and the standard deviation 

associated with each mean value should decrease. Depending on the code used by the 

applicant, a number of diagnostic calculations are generally available to demonstrate adequate 

convergence and statistical variation. For deterministic codes, a convergence limit is often 

prescribed in the input. The selection of a proper convergence limit and the achievement of this 

limit should be described and demonstrated in either the SAR or supporting criticality 

calculations. When burnup credit is included in the criticality analysis, the reviewer needs to be 

sure that proper neutron sampling and convergence have been achieved because the flux will 

be concentrated in the low burned ends of the fuel assemblies. 

(HIGH Priority) Because of the importance and complexity of the criticality evaluation, 

independent calculations should be performed to ensure that the most reactive conditions have 

been addressed, the reported keff is conservative and the applicant has appropriately modeled 

the storage cask geometry and materials. In deciding the level of effort necessary to perform 

independent confirmatory calculations, the reviewer should consider the following factors: 

(1) the calculation method (computer code) used by the applicant, (2) uniqueness and 

complexity of the design and analysis, (3) the degree of conservatism in the applicant’s 

assumptions and analyses, and (4) the extent of the margin between the calculated result and 

the acceptance criterion of keff < 0.95. As with any design and review, a small margin below the 

acceptance criterion and/or a small degree of conservatism may necessitate a more extensive 

staff analysis. 
 7-11 

(HIGH Priority) The reviewer should develop a model that is independent of the applicant’s 

model. If the reported keff for the most reactive case is substantially lower than the acceptance 

criterion of 0.95, a simple model known to produce very bounding results may be all that is 

necessary for the independent calculations. 

(HIGH Priority) If possible and appropriate, the reviewer should perform the independent 

calculations with a computer code different from that used by the applicant. Likewise, use of a 

different cross-section set, derived from a different cross-section database where possible and 

appropriate (e.g., ENDF/B, JEF, JENDL, UKNDL, etc.), can provide a more independent 

confirmation. The continuous energy (CE) cross sections created for use with KENO in the 

SCALE code system are generated by the AMPX processing code rather than the more widely 

used NJOY code. Even though some cross section libraries may not have fully independent 

data bases because they are all derived from ENDF/B data, the CE library in SCALE still can 

provide some level of independence and is useful for checking computations performed with 

libraries which were generated by using NJOY. The reviewer should describe the staff’s 

independent analysis and the analysis general results and conclusions in the SER. 

(HIGH Priority) Although a keff of 0.95 or lower meets the acceptance criterion, the reviewer 

should watch for design features or content specifications where small changes could result in 

large changes in the value of keff. When the value of keff is highly sensitive to system 

parameters that could vary, the acceptable keff limit may need to be reduced below 0.95. When 

establishing a keff limit below 0.95, the reviewer should consider the degree of sensitivity to 

system parameter changes and the likelihood and extent of potential parameter variations. 

7.5.4.3 Benchmark Comparisons (HIGH Priority) 

Computer codes for criticality calculations should be benchmarked against critical experiments. 

A thorough comparison provides justification for the validity of the computer code, its use for a 

specific hardware configuration, the neutron cross sections used in the analysis, and 

consistency in modeling by the analyst. Ultimately the benchmarking process establishes a bias 

and uncertainty for the particular application of the code (using the benchmark results for 

calculations performed by another analyst does not address this last issue) . The calculated keff

of the cask should then be adjusted to include the appropriate biases and uncertainties from the 

benchmark calculations. 

The reviewer should examine the general description of the benchmark comparisons. This 

examination includes verifying that the analysis of the experiments used the same computer 

code, computer system, cross-section data, modeling methods, and code options that were 

used to calculate the cask system keff values. 

 

The reviewer should also closely examine the applicant’s benchmark analysis to determine 

whether the benchmark experiments are relevant to the actual cask design. No critical 

benchmark experiment will precisely match the fissile material, moderation, neutron poisoning, 

and configuration in the actual cask. However, the applicant can perform a proper benchmark 

analysis by selecting experiments that adequately represent cask and fuel features and 

parameters that are important to reactivity. Key features and parameters that should be 

considered in selecting appropriate critical experiments include the type of fuel, enrichment, 

hydrogen-to-uranium (H/U) ratio (dependent largely on rod diameter and pitch), reflector 

material, neutron energy spectrum, and poisoning material and placement. The applicant 

should justify, and the reviewer should verify, the suitability of the critical experiments chosen to 

benchmark the criticality code and calculations. Techniques such as the sensitivity/uncertainty 
 7-12 

method developed by Oak Ridge National Laboratory (ORNL/TM-2005/39, 2005) can be helpful 

when assessing the applicability of the critical experiments used to benchmark the design 

analysis. UCID-21830 (Lloyd, 1990), the “International Handbook on Evaluated Criticality 

Safety Benchmark Experiments,” (NSC,NEA, 9/2003) and NUREG/CR-6361 provide information 

on benchmark experiments that may apply to the cask being analyzed. 

The reviewer needs to assess whether the applicant analyzed a sufficient number of appropriate 

benchmark experiments and how the results of these benchmark calculations have been 

converted to a bias for the cask calculations. Simply averaging the biases from a number of 

benchmark calculations typically is not sufficient, such as when one benchmark yields results 

that are significantly different from the others, the number of experiments is limited, or 

benchmarks that over-predict keff are included. In addition, benchmark comparisons should be 

checked for bias trends with respect to parameter variations (such as pitch-to-rod-diameter 

ratio, assembly separation, reflector material, neutron absorber material, etc.). A Lawrence 

Livermore National Laboratory (LLNL) (Lloyd, 1990) and NUREG/CR-6361 provide some 

guidance, but other methods, when adequately explained, have also been considered 

appropriate. 

For Monte Carlo codes, the statistical uncertainties of both benchmark and cask calculations 

also need to be addressed. The uncertainties should be applied to at least the 95-percent 

confidence level. As a general rule, if the acceptability of the result depends on these rather 

small differences, the reviewer should question the overall degree of conservatism of the 

calculations. Considering the current availability of computer resources, a sufficient number of 

neutron histories can readily be used so that the treatment of these uncertainties should not 

significantly affect the results. 

The reviewer should verify that only biases that increase keff have been applied. For example, if 

the benchmark calculation for a critical experiment results in a neutron multiplication that is 

greater than unity, it should not be used in a manner that would reduce the keff calculated for the 

cask. Only corrections that increase keff should be applied to preserve conservatism. 

The reviewer may have already performed a number of benchmark calculations applicable to 

storage casks and may have a reasonable estimation of the bias to be applied to the 

independent calculation of the cask. If such is not the case, or if the acceptability depends on 

small bias differences, the reviewer again needs to determine whether sufficient conservatism 

has been applied to the calculations. 

7.5.5 Burnup Credit (HIGH Priority) 

For guidance regarding the use of burnup credit, see the current revision of ISG-8. 

7.5.6 Supplemental Information 

The reviewer should ensure that all supportive information or documentation is provided. This 

may include, but not be limited to, justification of assumptions or analytical procedures, test 

results, photographs, computer program descriptions, input/output, and applicable pages from 

referenced documents. In addition, the SAR should include a list of fuel designs with the 

acceptable parametric limits and the maximum enrichments for which the criticality analysis is 

valid. The reviewer should request any additional information needed to complete the review. 
 7-13 

7.6 Evaluation Findings 

The reviewer should review the 10 CFR Part 72 acceptance criteria and provide a summary 

statement for each. These statements should be substantially as follows: 

 F7.1 Structures, systems, and components important to criticality safety are described 

in sufficient detail in Chapters of the SAR to enable an evaluation of their 

effectiveness. 

 F7.2 The cask and its spent fuel transfer systems are designed to be 

subcritical under all credible conditions. 

 F7.3 The criticality design is based on favorable geometry, fixed neutron poisons, and 

soluble poisons of the spent fuel pool [as applicable]. An appraisal of the fixed 

neutron poisons has shown that they will remain effective for the term requested 

in the CoC application and there is no credible way for the fixed neutron poisons 

to significantly degrade during the requested term in the CoC application; 

therefore, there is no need to provide a positive means to verify their continued 

efficacy as required by 10 CFR 72.124(b). 

 F7.4 The analysis and evaluation of the criticality design and performance have 

demonstrated that the cask will enable the storage of spent fuel for the term 

requested in the CoC application. 

The reviewer should provide a summary statement similar to the following: 

“The staff concludes that the criticality design features for the [cask designation] are in 

compliance with 10 CFR Part 72, as exempted [if applicable], and that the applicable 

design and acceptance criteria have been satisfied. The evaluation of the criticality 

design provides reasonable assurance that the [cask designation] will allow safe storage 

of spent fuel. This finding is reached on the basis of a review that considered the 

regulation itself, appropriate regulatory guides, applicable codes and standards, and 

accepted engineering practices.” 
 8-1 

8 MATERIALS EVALUATION

8.1 Review Objective 

The materials review ensures adequate material performance of components important to 

safety of a dry cask storage system (DSS), including the spent fuel canister or cask, under 

normal, off-normal, and accident-level conditions. To ensure an adequate margin of safety in 

the design basis of the DSS, the reviewer should obtain reasonable assurance that: 

 • The physical, chemical, and mechanical properties of materials for components 

important to safety (ITS) meet their service requirements including normal, offnormal, and accident-level conditions, and that the mechanical properties are 

Code accepted values. 

 • Materials for components ITS have sufficient requirements to control the quality 

of the production, fabrication, and test activities. 

 • Materials for ITS components are selected to accommodate the effects of, and to 

be compatible with, the independent spent fuel storage installation (ISFSI) site 

characteristics, environmental conditions, and duration of the license period. 

 • The spent nuclear fuel (SNF) cladding is protected from gross rupture and from 

conditions that could lead to fuel redistribution. 

 • The DSS is designed to maintain the spent fuel in a readily retrievable condition. 

 • Other materials which support or protect ITS components (such as coatings) are 

suitable for the application. 

In reviewing the materials, the reviewer should consider the sources of information for the 

physical and mechanical properties of the materials used in the DSS construction and those 

materials which are part of the spent fuel payload. These material properties should be 

considered against both static and dynamic loadings for normal, off-normal, accident conditions, 

and other phenomena such as corrosion. The material properties and characteristics needed to 

satisfy these functional safety requirements should be maintained and are applicable over the 

complete licensing period. 

Preferred materials information sources are U.S. industry consensus codes, standards, and 

specifications. The applicability and acceptability of all other sources, such as manufacturer’s 

test data and handbooks, should be reviewed. The reviewer should also examine published 

articles, research reports, and texts as sources of information concerning material performance. 

Foreign standards (and codes) may be acceptable on a case by case basis. The applicant 

should provide complete documentation supporting the use of the foreign standard and show 

that the foreign standard is equivalent to a comparable US standard (e.g. ASME, ASTM, etc.), 

or otherwise sufficient for its intended use. The staff may need to review foreign standards in 

greater depth, depending on the familiarity with the standard and applicability of the standard to 

the proposed DSS design 
 8-2 

8.2 Areas of Review 

The materials evaluation encompasses the following listed areas of review. The various 

materials engineering related topics requiring review may be addressed in different chapters of 

the SAR. However, the review guidance for all materials engineering related topics are 

provided in this chapter of the SRP. 

Areas for materials review: 

General 

 Cask Design/Materials 

 Environmental Conditions 

Engineering Drawings 

Materials Selection

 Applicable Codes and Standards and Alternatives to the Code 

Material Properties 

 Alternative or Substitute Materials (ITS components) 

 Copper bearing or other weathering steels or other corrosion control measures 

 for coastal ISFSI locations 

Weld Design, Inspection 

Bolt Applications 

Coatings 

Neutron Shielding Materials 

 Gamma shielding 

Neutron Poison Materials for Criticality Control 

Concrete and Reinforcing Steel 

Seals 

 Low Temperature Ductility of Ferritic Steels 

 Creep Properties/Analyses 

 

Corrosion

 Corrosion Resistance 

 Galvanic/Chemical/Radiolytic Reactions of Fuel with Canister Internals 

 

Cladding Integrity/Fuel

 Fuel Burn-up 

 Cladding Temperature Limits 

Damaged Fuel Definition 

Operational Issues (see Operating Procedures Chapter of SAR) 

 Hydrogen gas monitoring/mitigation 

Preventing oxidation of fuel during loading/unloading operations which can lead 

to Rod Splitting 
 8-3 

Examination and Testing (see Acceptance Test Chapter of SAR) 

 Helium leakage testing of canister welds 

 Periodic Inspections 

Code Case Acceptability

 Refer to Regulatory Guide 1.193 

8.3 Regulatory Requirements 

This section presents a summary matrix of the portions of U.S. Code of Federal Regulations 

(CFR) Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and 

High-Level Radioactive Waste,” Title 10, “Energy” (10 CFR Part 72) relevant to the review areas 

addressed by this chapter. The U.S. Nuclear Regulatory Commission (NRC) staff reviewer 

should read the exact referenced regulatory language. Table 8-1 matches the relevant 

regulatory requirements associated with this chapter to the areas of review. 

Table 8-1 Relationship of 10 CFR Part 72 Regulations and Areas of Review 

Chapter 8 

Areas of Review

10 CFR Part 72 Regulations

72.104(a) 72.106(b) 

72.122 

(a), (b), 

(c) 

72.122 

(h)(1), (i), 

(l) 

72.124 

General 

Materials Selection ● ● ● ●

Corrosive Reactions 

Cladding Integrity ● 

Chapter 8 

Areas of Review

10 CFR Part 72 Regulations

72.236(g) 72.236(h) 72.236(i) 72.236(m) 

General ●

Materials Selection ● ● ●

Corrosive Reactions ● 

Cladding Integrity ●

8.4 Review Procedures and Acceptance Criteria 

Metallic materials are primarily assumed in this guidance. The interrelationship of the materials 

evaluation review with other disciplines is shown in Figure 8-1. 

8.4.1 General Review Considerations (HIGH Priority) 

The reviewer should survey the SAR and design drawings (generally SAR Chapters 1 and 2) to 

identify the various materials issues that may be associated with the specific design proposal in 
 8-4 

the application. The reviewer should also examine the criticality, shielding, confinement, and 

thermal chapters to identify cross-cutting issues that should be coordinated among the technical 

disciplines. Note, not all design and license changes in the amendment will necessarily be 

neither separately identified by the applicant nor obvious. 

The reviewer should examine the following Technical Specification (TS) items to verify its 

proposal by the applicant and understand the specific limits, design requirements, and operating 

constraints proposed by the applicant: 

 Maximum fuel burn-up 

 Maximum cladding temperature 

 Definition of damaged fuel 

 Code of record and alternatives to specific Code requirements 

 Specification/requirements for alternative materials for ITS components 

 Manufacture and testing of neutron poison material(s) for criticality control 

 Hydrogen monitoring/mitigation during wet loading/unloading 

 Helium leakage testing of confinement and cover welds 

 Maintaining inert atmosphere during canister draining/flooding to prevent oxidation 

 Use of Code Case N-595 (not acceptable) 

 Use of copper bearing or weathering steel for structural steel components at coastal 

marine ISFSI sites (or other corrosion mitigation measures) 

 Operational controls to maintain cladding temperature limits 

 Low Temperature Ductility of Ferritic Steels 

 Damaged fuel definitions 

 Materials acceptance testing 

Design temperature for aluminum components used in the fuel basket or canister interior 

(creep issues) 
 8-5 

 

Figure 8-1 Overview of Materials Evaluation 

Chapter 1 – General 

Information Evaluation 

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation 

• DSS Description and 

Operational Features 

• DSS Drawings and 

Design Information 

• Structural and Pressure 

Testing 

• Material Fabrication 

Verification 

• Maintenance Program 

Chapter 12 – Accident 

Analysis Evaluation 

• Material Properties 

• Corrosive Reactions 

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation 

• Operating Monitoring and 

Safety Limits 

• Loading, Unloading, and 

Handling Limits 

Chapter 3 – Structural 

Evaluation 

• Applicable Codes and 

Standards 

• Material Properties 

• Weld Design and 

Specification 

• Bolt Applications 

Chapter 5 – Confinement 

Evaluation 

• Closure Specifications 

• Weld Design and 

Specifications 

• Bolt Applications 

• Seals 

Chapter 2 – Principal 

Design Criteria Evaluation 

• Design Basis for SSCs 

Important to Safety 

• Design Criteria for Safety 

Protection Systems 

• Material Compatibility 

Chapter 4 – Thermal 

Evaluation 

• Component Materials 

• Heat Transfer Properties 

• Cask Drying 

• Temperature Limits 

Chapter 5 – Confinement 

Evaluation 

• Confinement Design 

• Seal Material 

• Weld Design 

Chapter 4 – Thermal 

Evaluation 

• Material Properties 

• Temperature Limit 

Chapter 6 – Shielding 

Evaluation 

• Properties of Shielding 

Material 

• Shielding Design Criteria 

Chapter 3 – Structural 

Evaluation 

• Structural Design Criteria 

• Structural Materials 

• Structural Compatibility 

with Functional 

Performance 

Chapter 7 – Criticality 

Evaluation 

• Material Properties 

• Fuel Specifications 

Corrosion Reactions 

• Loss of Corrosion Resistance 

• Gas Generation 

Cladding Integrity 

• Temperature Limits 

• Oxide Thickness 

• Damaged Fuel 

Appendix 

• Technical Basis for 

Temperature Limits 

• Computation of Percent Credit 

for Boron-Based Neutron 

Absorbers 

Materials Selection 

• Applicable Codes and Standards and 

Alternatives to the Code 

• Material Properties 

• Weld Design and Specification 

• Bolt Applications 

• Coatings 

• Gamma and Neutron Shielding Materials 

• Neutron Absorbing/Poison Materials for 

Criticality Control 

• Concrete and Reinforcing Steel 

• Seals 

General 

• Cask Design/Materials, Fuel 

Specifications, and 

Environmental Conditions 

• Engineering Drawings 

Chapter 8 – Materials Evaluation 

Chapter 9 – Operating 

Procedures Evaluation 

• Thermal Cycling 

• Gas Generation 

Detection 

• Closure 
 8-6 

8.4.2 Codes and Standards (HIGH Priority) 

8.4.2.1 Usage and Endorsement 

Codes (or “construction codes”) govern which materials may be used and how they may be 

employed. Standards detail how a material is produced and establishes chemical and material 

property requirements. All ASME materials are a subset of AWS and ASTM materials. 

However, not all ASTM materials are endorsed for use by the ASME or other codes which may 

be used for canister design. 

 

The SAR must identify applicable codes and standards used in the design, selection, and use of 

materials. For important-to-safety (ITS) components, U.S. industry consensus codes and 

standards such as ASME, AWS, ANSI, ACI, and ASTM should be specified. 

Foreign codes and standards are generally NOT acceptable for ITS components/materials and 

would only be approved on a case-by-case basis. However, foreign-produced materials which 

comply with U.S. codes and standards are acceptable. 

ITS components subject to ASME Section III jurisdiction, typically confinement boundary and 

fuel basket, are normally ASME Section II materials. ITS attachments to the confinement 

boundary, as well as structural components of the overpack, may be ASME or ASTM materials, 

depending on the code of record for the component. For non-ASME ITS components, ASTM 

materials may be used. 

Non-ITS items can be specified by generic names such as “stainless steel”, “aluminum,” “carbon 

steel,” etc., as appropriate for the application. 

Proprietary materials which are ITS (specifically neutron poisons) must be described adequately 

in SAR Chapter 8, “Materials” to permit the staff to make a safety finding. The governing quality 

assurance and quality control (QA/QC) documents, key manufacturing procedures, and key 

testing protocols for proprietary materials should be incorporated by reference into the TS. 

Limited changes to the materials composition, performance, or manufacturing methods may be 

allowed if the changes satisfy the criteria of 10 CFR 72.48. 

Polymeric neutron shielding materials, which are usually proprietary, are considered importantto-safety (ITS) materials in order to meet the regulatory requirements of 72.126(a)(6). 

NUREG/CR-6407 specifically designates neutron shielding materials as ITS Category B. 

8.4.2.2 Code Case Use/Acceptability 

Review any referenced ASME Code cases against Regulatory Guide 1.193 for acceptability. 

Note that Code Case N-595 (any revision) has been found unacceptable to the staff per 

RG 1.193. 

8.4.3 Environment (Priority – as indicated) 

(MEDIUM Priority) Generally, the ISFSI site with associated storage canisters are subjected 

(long-term) to a mild atmospheric environment. Twenty or more years of ISFSI operational 

experience has verified that no significant corrosion issues generally exist during storage. 

However, note whether or not the site or potential site is a coastal marine location. Additional 

corrosion prevention measures may be applied when the ISFSI is located in a coastal marine 
 8-7 

environment. Detailed review guidance is provided in 8.4.6 Coastal Marine ISFSI Sites–

Material Selections. 

(LOW Priority) Underground structures require additional consideration due to soil corrosion 

issues. Additional guidance is provided in 8.4.14.3 Omission of Reinforcement. 

(LOW Priority) Fuel loading/unloading conditions assume a borated, demineralized water 

environment at temperatures up to the boiling point. Experience with the conventional stainless 

steel and aluminum construction canister internals have verified no significant corrosion of fuel 

canister ITS components occur during the limited duration of a fuel loading/unloading operation. 

Pool water is buffered to a pH of about 8.5 to limit corrosion. 

8.4.4 Drawings (MEDIUM Priority) 

Licensing drawings usually appear in SAR Chapters 1 or 2. Examine the drawings and drawing 

notes for material specifications and alternatives. Ensure any materials substitutes are 

adequately specified, either on the drawing or in the SAR. ITS component material substitutes 

must appear in the TS. 

8.4.5 Material Properties (MEDIUM Priority) 

8.4.5.1 Structural Properties 

The intent of this portion of the materials evaluation is to determine the acceptability of all 

material properties that have a structural role in confinement system structures and other 

structures important to safety (e.g., the basket, impact limiters, and shielding) and non-safety. 

The material properties and characteristics need to be applicable over the term requested in the 

CoC application. The reviewer should analyze the potential for corrosion and ensure that the 

applicant established and used appropriate corrosion allowances for the structural analyses. 

The range of some materials components properties may have to be evaluated over the range 

of life cycle conditions experienced during cask fabrication, loading, emplacement, storage, 

transfer, retrieval, unloading, and decontamination. 

The information provided on structural materials must be consistent with the application of 

accepted design criteria, codes, standards, and specifications selected for the storage cask 

system and as described in this chapter and Chapter 3, “Structural Evaluation” of this SRP. 

Materials and material properties used for the design and construction of these safety-related 

structures should comply with the applicable codes and standards identified in Section 

3.5.2.2 (i). For example, if the applicant elects to use design criteria from Section III of the 

ASME B&PV Code, the materials selected for the cask must be consistent with those allowed 

by the ASME Code subsection related to design. Acceptable requirements include the ASME 

adopted specifications given in Section II, Part A, “Ferrous Metals;” Part B, “Nonferrous Metals;” 

Part C, “Welding Rods, Electrodes, and Filler Metals;” and Part D, “Properties.” The review of 

structural materials should be coordinated with the structural discipline. 

A list of all materials used and the proposed service conditions for those materials during 

loading, storage, and unloading is a useful aid during the review. These tables provide various 

types of information that the reviewer needs from an application to aid in determining the 

suitability of the materials for the structural evaluation. The tables include the name and safety 

classification of each component part of the DSS and, where applicable, the function, the 

material specification(s) to which it is produced, and the nominal values for structural 
 8-8 

parameters. The tabulation should include all materials used for components with an importantto-safety function (e.g., confinement, transfer, criticality control, shielding). Information in this 

table can aid the reviewer to formulate the types of performance-related questions that are 

important for each component of a storage system. 

The SAR documentation should fully define the structural materials used for components 

important to safety. The reviewer may find it useful to tabulate the major structural materials to 

facilitate the review. The following information could be tabulated: specification number, grade, 

type, and class of the material, nominal composition, product form, yield strength, tensile 

strength, and notes about the materials, etc. The SAR should identify properties related to 

structural performance and resistance or response to thermal, radiation, or other applicable 

environments that may impact structural performance. The structural and material disciplines 

should coordinate their reviews as appropriate for these components. 

The completeness, accuracy, and acceptability of the identification and stated properties of the 

safety-related materials should be reviewed. In reviewing the structural materials, the reviewer 

should consider the sources of information; properties used in the structural evaluation and 

suitability for term requested in the CoC application. The reviewer should verify that the SAR 

clearly references acceptable sources of all material properties. 

Examine the SAR adopted material properties for ITS component materials and ensure ASME 

Section II, Part D, properties and stresses are employed. The longstanding staff position 

(developed by NRR) regarding material properties is that ASME Code values must be used. 

Use of certified material test report (CMTR) values of UTS, yield, etc., is generally not 

permissible. Use of CMTR values is at risk of being non-conservative because samples may be 

taken at a portion of the ingot, billet, or forging that have optimum materials properties during 

certification. 

.

8.4.5.2 Thermal Materials 

The materials reviewer should coordinate with the thermal reviewer to determine the materials 

properties of the materials important to the thermal analysis. The material compositions and 

thermal properties such as thermal conductivity, thermal expansion, specific heat, and heat 

capacity should be verified as a function of the temperature over the range the components are 

to operate, for all components used in the safety analysis. Verify the change in these material 

properties due to potential degradation of materials over their service life has been evaluated by 

the applicant. Temperature and anisotropic dependencies of thermal properties should be 

considered.

 8.4.6 Coastal Marine ISFSI Sites–Material Selections (MEDIUM Priority) 

At coastal marine locations, the heavy salt drift can significantly accelerate the normally slight 

atmospheric corrosion rate to unacceptable values of some canister storage module designs, 

such as those that employ carbon steel structural elements inside the canister storage module. 

Experience has shown ordinary grades of structural steel (such as A-36) withstand the 

nominally dry interior environment of the canister overpack very well over a 20 year operational 

period. 

For such cases, the reviewer must verify that the corrosion allowance specified is adequate for 

the 20 to 40 year CoC period of the canister. Corrosion rates for carbon steel in air may be 

found in corrosion references such as Corrosion Engineering by Fontana and Greene, 
 8-9 

Corrosion Data Survey by the National Association of Corrosion Engineers (NACE), Corrosion 

and Corrosion Control by Uhlig, and the publications of the NASA Kennedy Space Center 

Corrosion Technology Laboratory. For exposures to coastal marine atmospheres, the corrosion 

rate data from the Kennedy Space Center Corrosion Technology Laboratory appears to be 

bounding for any location in the continental United States. 

To address the increased atmospheric corrosion rates found at coastal marine (salt water) sites, 

some applicants have specified the use of 0.20%, minimum, copper-bearing steels, or, 

“weathering steels” such as Cor-Ten. The Kennedy Space Flight Center has collected data 

which has demonstrated the benefit of copper-bearing and weathering steels for significantly 

reducing corrosion at coastal marine sites. Therefore, for coastal marine ISFSI sites, the use of 

copper-bearing steels (containing a minimum of 0.20 percent copper), or weathering steels, may 

be necessary. Such steels are covered by ASTM A-242 and A-588, and supplemental 

requirements to ASTM A-36, and/or other specifications. 

Other corrosion control measures may be employed, provided adequate documentation is 

supplied to demonstrate efficacy. 

Coatings may be specified to alleviate the coastal atmospheric corrosion issue. However, 

unless supporting data is available to demonstrate the predicted coating life, the coating must 

be periodically inspected and maintained. 

8.4.7 Weld Design/Inspection (MEDIUM Priority) 

8.4.7.1 Welding Codes–Background Discussion 

The nationally recognized codes which have been used for spent fuel canister construction 

include: 

 • ASME B&PV Code, Section III, “Rules for Construction of Nuclear Facility 

Components,” Division 1. 

 • AWS D1.1 (current edition), “Structural Welding Code-Steel.” 

• AWS D1.6 (current edition), “Structural Welding Code-Stainless Steel.” 

The ASME B&PV Code Section III contains the design requirements for nuclear systems at a 

commercial nuclear power plant. It contains sections governing the design of welded nuclear 

components in the plant. 

AWS D1.1 is the structural welding code for carbon steel structures such as bridges and steelframed buildings. 

The NRC staff accepts the use of the ASME B&PV Code, Section III, as the preferred 

construction code for storage casks. Some older cask designs used the AWS D1.1 Code. 

Note, the various construction codes (e.g., ASME Sections I, III, or VIII, and AWS D1.1) differ 

from one another in their requirements for materials and welding procedures, because each 

code is specialized with a particular application in mind. 

The ASME construction codes are supplemented by “supporting codes” which detail how 

special processes such as welding and nondestructive examination (NDE) are to be qualified 
 8-10 

and executed. ASME B&PV Code Section IX, “Welding and Brazing Qualifications” details the 

requirements for specifying and qualifying a welding procedure and for testing and qualifying 

welders. ASME B&PV Code Section V, “Nondestructive Examination,” supports the various 

ASME construction codes by detailing the required qualifications for NDE examiners and the 

requirements and methods for performing the types of NDE specified by the various 

construction codes. 

Standard welding and NDE symbols may be found in AWS A2.4 (latest edition), “Symbols for 

Welding, Brazing, and Nondestructive Testing,” to aid interpretation of such symbols found on 

the drawings submitted with the SAR. 

Technical specification items related to the welds and testing are discussed separately. 

8.4.7.2 Weld Design and Testing 

Verify that the canister confinement welds are full penetration welds. Inspection of these welds 

must follow the ASME Code requirements of full volumetric examination [radiographic testing 

(RT) or ultrasonic testing (UT)] and a surface examination [liquid penetrant testing (PT), for 

austenitic stainless steel canisters]. A hydrostatic or pneumatic test is also required by the 

Code. 

Stainless steel fillet welds can only be inspected by PT. Volumetric inspection of fillet welds is 

not feasible. 

Due to the relatively benign operating conditions in storage, imposition of specific weld filler 

metals, or use/prohibition of certain welding processes is not presently necessary. Sensitization 

of the stainless steel is not an issue. Hence, solution annealing is unnecessary. 

A shop helium leakage test, using ANSI N14.5 testing standards, must be performed to 

demonstrate that the entire canister or cask confinement body is free of defects that could lead 

to a leakage rate greater than the allowable design basis leakage rate specified in the 

confinement analyses. The requirements for the helium leakage test should be specified in the 

CoC to meet the requirements of 10 CFR 72.236(j) and (l). For bolted closure casks the entire 

confinement boundary should be similarly helium leak tested and pressure tested. The 

confinement boundary should be tested at the fabrication shop, with only a leakage test 

performed on the bolted lid closure seals (including drain and vent port seals) tested in-field by 

the cask user. The lid-to-shell welds and vent ports should be fabricated and helium leakage 

tested in accordance with the guidance of Section 8.4.20, as applicable. The staff should note 

that only lid-to-shell welds are within the scope of leak testing exceptions specified in 8.4.20. 

The entire confinement boundary should be pressure tested hydrostatically or pneumatically to 

125 or 110 percent of the design pressure, respectively. The test pressure should be 

maintained for a minimum of 10 minutes prior to initiation of a visual examination for leakage, 

per the ASME Code. 

Following the application of the test pressure for the required time, all joints, connections, and 

regions of high stress, such as regions around openings and thickness transition sections, 

should be visually examined for leakage. This visual examination shall be performed in 

accordance with ASME Code requirements and shall be performed at a pressure equal to or 

greater than the design pressure or three-fourths of the test pressure. This pressure test and 
 8-11 

visual examination applies to both the canister body constructed at a fabrication facility and the 

lid-to-shell welds fabricated and closed in the field by a Part 72 licensee. 

If pressure testing is performed only in the field, the visual examination of the portions of the 

canister shell may be impractical due to its inaccessibility inside the transfer cask. The 

application should discuss the proposed operations and reasons for inaccessibility for visual 

examination. Due to the inability to perform the visual examination of inaccessible portions of 

the canister welds during the field ASME Code hydrostatic test, staff has accepted the results 

from the shop helium leakage test applied under ANSI-N14.5 standards. The exception and 

basis should be listed in the table of ASME code exceptions in the Certificate of Compliance 

(CoC). 

After the canister is loaded and lids welded, the confinement welds are pressure tested and 

helium leakage rate tested as further detailed in section 8.4.20. 

8.4.7.3 Lid Welds and Closure Welds 

The staff should verify the cask design is in compliance with Section 8.9 of this SRP or as 

follows: 

 • This guidance only applies to canisters of all-welded construction, fabricated from 

austenitic stainless steel, and employing redundant welds for the confinement 

closure. 

• The welded canister (i.e., the confinement boundary) must be leak tested in 

accordance with ANSI N14.5-1997, except as specified by this guidance. The 

exemption for leak testing only applies to the closure welds that are typically 

made in the field and all other welds should be leak tested. 

 • “Structures, systems, and components important to safety must be designed to 

withstand postulated accidents” (10 CFR 72.122(b)). 

 • Records documenting the lid welds shall comply with the provisions of 10 CFR 

Part 72.174, “Quality Assurance Records” or with NQA-1, “Quality Assurance 

Requirements for Nuclear Facility Applications,” depending upon the standard in 

effect at the time of licensing. 

• Activities related to inspection, evaluation, documentation of fabrication, and lid 

welding shall be performed in accordance with an NRC-approved quality 

assurance program as required in 10 CFR Part 72, Subpart G, “Quality 

Assurance.” 

A redundant sealing of the canister is required by 10 CFR 72.236(e). One of the redundant 

seals in a welded canister design will involve a structural weld. The structural lid weld joint will 

be a full or partial penetration groove weld. 

Carbon and Alloy Steel Cask Designs

The reviewer should verify the applicant has considered all the closure lid weld material and 

technique improvements that accrued from previous DSS design and fabrication experience. 

For example, the reviewer should refer to the technical evaluation in NRC Confirmatory Action 
 8-12 

Letter 97-7-001, 1998 (ADAMS ML 060620420). Some of the DSS improvements resulting from 

that action include: 

 • Shell plates made from low sulfur, calcium-treated, vacuum-degassed steel. 

 • Application of minimum 93°C (200°F) preheat. 

 • Use of low-hydrogen electrodes. 

 • Low carbon equivalent base metals and weld metals. 

 • Magnetic particle examination (MT) of the root pass. 

 • Maintenance of preheat as a postheat treatment for a minimum of one hour. 

 • Minimum of two-hour delay after postheat before performing final volumetric 

NDE. 

UT examine the structural lid weld in accordance with ASME Section III, D1, NB method and 

acceptance criteria requirements 

Progressive surface examinations, utilizing a PT or magnetic particle testing (MT), are permitted 

only if unusual design and loading conditions exist. In addition, a stress-reduction factor of 0.8 

is imposed on the weld strength of the closure joint to account for imperfections or flaws that 

may have been missed by progressive surface examinations. The weld design should be 

approved by the NRC on a case-by-case basis. 
 8-13 

8.4.7.4 Austenitic Stainless and Nickel-Base Alloy Steels Cask Design 

NDE of the large structural lid-to-shell weld designs fabricated from austenitic materials may be 

volumetric UT or multi-pass PT examined. A multi-pass PT is defined as performing a PT 

inspection of every pre-calculated intermediate weld deposit depth (layer) between the root and 

final weld layers. 

Use ASME Section III, Division 1, Subsection NB (Section III, D1, NB) requirements for UT and 

PT inspection method and acceptance criteria. 

A multiple-pass PT examination may be utilized in lieu of UT inspection and is performed as 

follows: Note: Impose a stress reduction factor of 0.8 for weld strength. 

1. Calculate the critical flaw size (depth) assuming a buried flaw. Postulate a full 

circumferential (360-degree) flaw. Use ASME Section XI, D1, IWB 3600 requirements 

for alternative flaw acceptance criteria. Use of J-integral or net section stress is 

acceptable. 

2. Establish the maximum allowable intermediate weld deposit depth (layer)/required inprocess PT inspection interval by using the critical flaw depth calculated in Step 1. Note: 

Lessons learned suggest that the critical flaw depth for many structural lid welds is 3/8-

inch. 

3. PT the root layer, every intermediate layer established in Step 2 and the final weld layer. 

It is assumed that the root layer is single pass. If the root layer is multi-pass, calculate 

the critical flaw depth (Step 1) to establish the maximum allowable intermediate weld 

deposit depth (layer)/required in-process PT interval. Assume a surface connected flaw 

when calculating the critical flaw depth for a multi-pass root layer. 

The applicant’s evaluation of the critical flaw size using the above methodology should be 

reviewed based on service temperature, dynamic fracture toughness and critical design stress 

parameters as specified in ASME Section XI, D1. 

8.4.8 Galvanic/Corrosive Reactions (LOW Priority) 

8.4.8.1 Environmental considerations 

The reviewer can find operational issues associated with hydrogen generation and guidance for 

evaluating galvanic or corrosive reactions in NRC Bulletin 96-04 (1996). The reviewer should 

confirm the DSS will perform adequately under the operating environments expected (e.g., 

short-term loading/unloading or long-term storage) for the duration of the license period such 

that no adverse galvanic or corrosive reactions occur between the canister materials, fuel 

payload, and the operating environments. 

8.4.8.2 Canister Contents 

The staff has previously reviewed a number of non-fuel hardware components and materials for 

compliance with 10 CFR 72.120(d), meaning, compatibility with a canister interior composed of 

stainless steel and aluminum components. These components are various neutron source 

assemblies, burnable poison rod assemblies (BPRAs), thimble plug devices, and other types of 
 8-14 

control elements. The staff has found the following materials to be acceptable for storage when 

the canister is constructed of stainless steel with stainless steel and aluminum basket 

components: 

 

Neutron source materials composed of stainless steel or zirconium alloy cladding containing: 

antimony-beryllium, americium-beryllium, plutonium-beryllium, polonium-beryllium, and 

californium. Exposure of these various contents to the wet loading and dry storage environment 

was assessed and found to be satisfactory. 

Control elements composed of zircaloy or stainless steel cladding containing: boron carbide, 

borosilicate glass, silver-indium-cadmium alloy, or thorium oxide. Exposure of these various 

contents to the wet loading and dry storage environment was assessed and found to be 

satisfactory. 

8.4.9 Creep Behavior of Aluminum Components (HIGH Priority) 

Aluminum based metal matrix composites and aluminum / boron carbide laminates (e.g. 

BoralTM) are employed for all presently utilized neutron poison materials. Also, aluminum 

components are frequently part of the spent fuel basket. More recent designs have specified 

ever higher design temperatures for the fuel basket components in order to accommodate 

higher loading densities and higher burn-up fuel. This trend has pushed the various aluminum 

components well into creep regime operating temperatures. 

Review the design maximum temperatures and stress for any aluminum components and verify 

a creep analysis has been performed if any structural load bearing aluminum components 

operate at a design temperature above approximately 200°F. 

 In the event temperatures exceed the ASME Section II nominal 400°F temperature limit for 

aluminum, other sources for creep data must be examined. One previously cited reference for 

this information is: D.W. Wilson, J.W. Freeman and H.R. Voorhees, Creep-Rupture Testing of 

Aluminum Alloys to 100,000 Hours, First Progress Report, prepared for the Metal Properties 

Council, New York, November 1969. The staff makes no judgment as to the acceptability of this 

reference. This is because the designs reviewed through the time of this writing have had 

design stresses (on the order of tens of PSI) which were substantially below the creep-rupture 

stresses provided in the referenced report. None-the-less, an assessment of creep deformation 

over a 20 to 40 year CoC period should be part of the design calculations. 

Borated aluminum neutron poison materials must be considered on a case-by-case basis if they 

are subjected to structural load bearing beyond their own dead-weight loads. This is due to 

their inherently low ductility and generally unknown creep properties. 

8.4.10 Bolt Applications (MEDIUM Priority) 

If threaded fasteners are employed for ITS components, verify the bolt material(s) have 

adequate resistance to corrosion and brittle fracture and a coefficient of thermal expansion 

similar to the materials being bolted together. 

8.4.11 Protective Coatings (LOW Priority) 

Coatings in DSSs are used primarily as corrosion barriers or to facilitate decontamination. They 

may have additional roles, such as improving the heat rejection capability by increasing the 

emissivity of cask internal components. Protective coatings are occasionally specified for 
 8-15 

carbon steel components. Coatings are not ITS components. The structures or components 

that the coatings are applied to are generally ITS component. No coating should be credited for 

protecting the substrate material or extending the useful life of the substrate material unless a 

periodic coating inspection and maintenance program is required for the coating. 

The staff has established this section to alleviate confusion regarding coatings on cask 

components. Coatings generally have low safety significance with the exception of coating 

issues that may result in adverse chemical or galvanic reactions. Typically, the detailed 

guidance in this section is not generally subject to further confirmation as part of the review. 

However, there may be instances in which unique or innovative coatings are specified by the 

applicant to perform a specific function unique to the cask system. In these instances, the 

reviewer may use discretion in implementing the detailed review guidance in this section. This 

section outlines methods and procedures for appropriately assessing coatings. Within the 

assessment several areas are covered in detail including the scope of the coating application, 

type of coating system, surface preparation methods, applicable coating repair techniques, and 

coatings qualification testing. 

8.4.11.1 Review Guidance 

The reviewer should determine the appropriateness of the coating(s) for the intended 

application by reviewing the coating specification for each protective coating that is applied to an 

important to safety component. A specification that describes the scope of the work, required 

materials, the coating’s purpose, and key coating procedures, should ensure that the 

appropriate and compatible coatings have been selected by the DSS designers. A coating 

specification should include the following: 

 • Scope of coating application; 

 • Type of coating system; 

 • Surface preparation methods; 

 • Coating application method; 

 • Applicable coating repair techniques; 

 • Coatings qualification testing, as applicable. 

8.4.11.2 Scope of Coating Application 

The coating specification should identify the purpose of the coating, a list of the components to 

be coated, and a description of the expected environmental conditions (e.g., expected 

conditions during loading, unloading, and dry storage). 

The reviewer should verify that the coatings will not react with the cask internal components and 

contents and will remain adherent and inert when exposed to the various environments of a 

SNF cask. The most prevalent, potentially degrading environments include the immersion in 

borated SNF pool water during loading and unloading operations, and high-temperature and 

high-radiation (including neutrons) environments encountered during vacuum drying evolutions 

and long-term storage.

8.4.11.3 Coating Selection 

The reviewer should verify that the coating specification identifies the manufacturer’s name, the 

type of primers and topcoat(s) comprising the coating system, and the minimum and maximum 

dry coating thickness(es). Due to the unique nature of coating properties, and coating 
 8-16 

application techniques, the manufacturer’s literature may be the only source of information on 

the particular coating. 

The reviewer should verify that the coating selected for cask components is capable of 

withstanding the intended service conditions over the design service life. Failures can be 

prevented by ensuring that the selection and the application of the coating are controlled by 

adhering to the coating manufacturer’s recommendations. 

8.4.11.4 Surface Preparation 

The reviewer should verify that the coating specification identifies whether solvent or abrasive 

cleaning methods should be used to prepare surfaces prior to coating application. This 

information should ensure that proper surface preparation techniques can be implemented 

during cask fabrication. 

The reviewer should confirm that the specified type and degree of surface cleaning and the 

required surface profile meet the coating manufacturer’s specification. Any deviations from the 

manufacturer’s standards for surface preparation must be supported by appropriate tests that 

demonstrate acceptable coating performance under all design conditions. 

8.4.11.5 Coating Repairs 

The reviewer should verify that the coating specification identifies the general requirements for 

repairing damage to the coating. This information will assist the reviewer in evaluating the 

effects of repairs on the integrity of the coating and whether the designated repair methods 

could be implemented during or after cask fabrication. 

The reviewer should examine the design to determine whether the structure is assembled 

before or after its various parts are coated. If a complex structure is to be coated after 

assembly, it is very important that the consequences of a potential coating failure be analyzed to 

determine whether other cask functions or component features could be compromised by the 

failure. 

The consequences of coating failure depend on the type of coating and service environment, 

and may include the following: 

 • Partial and/or complete coating failure that alters the corrosion resistance of DSS 

structural and shielding components (primarily during loading/unloading 

operations). 

 • Partial and/or complete coating failure that alters the emissivity and heat transfer 

of basket components. 

 • Particulates (cloudiness) that form in SNF pool water or cask during loading or 

unloading that may affect such operations. 

 • Aggressive or reactive chemical species that form and consequently impact the 

performance of other cask components during long-term exposure to radiation 

(e.g., gamma and neutron). 
 8-17 

8.4.11.6 Coating Qualification Testing 

Coatings used on cask external surfaces may have been selected upon the basis of their 

performance requirements and exposure conditions. The applicant may have used related 

industrial conditions as a documented guide or basis for coating selection without performing 

further laboratory tests. 

Any coating (including paints or plating) used inside a DSS must have been tested to 

demonstrate the coatings performance under all conditions of loading and storage. The 

conditions evaluated should include exposure to radiation, high temperature during vacuum 

drying and storage, and immersion during loading, unloading and transfer operations. The 

coating must be demonstrated to remain intact and inert for the full duration of the DSS design 

life. 

There are a number of standardized ASTM tests for coatings performance. In reviewing ASTM 

(or other) tests used to qualify coatings for service in storage casks, consideration should be 

given to the applicability of a test to the service conditions. 

Planning, execution, and interpretation of coating qualification tests must be performed by a 

qualified coatings engineer (e.g., certified by the National Association of Corrosion Engineers). 

The reviewer should ensure that appropriate, qualified expertise has been employed by the 

applicant for any coatings qualification program. 

The reviewer should verify that the coating specification includes a description of the coating 

qualifications testing program, as applicable. The following information, which is important to 

qualifying a coating, includes, but is not limited to: 

 • The size and shape of samples used for the coating tests, as well as the type of 

material(s), and a description and results of any tests conducted on partial or fullsize production mock-ups. 

 • The test sample surface preparation method(s) and expected or measured 

surface profile. Sample surface preparation should be performed in accordance 

with written production procedures, using the same equipment, materials, and 

qualified personnel as intended for production coating. Inspection methods and 

acceptance criteria should be included. 

 • Application method(s) and measured control parameters, including records of 

temperature and humidity, cure cycle and times, and any other monitoring or 

acceptance tests such as dry film thickness, hardness, and adhesion. The 

methods and parameters should be employed in accordance with written 

production procedures using the same equipment, methods, materials, and 

qualified personnel. 

 • A test plan description which clearly describes the rationale for and the types and 

sequences of all coating qualification tests, lab protocols, numbers of samples, 

inspection methods, and acceptance criteria. Raw test results should be 

tabulated or otherwise presented. The test plan should include: (1) laboratory 

coupons for demonstrating coating suitability/qualification; and (2) partial or full 

size production mock-up tests that demonstrate that the selected coating can be 

applied successfully to real production parts under production shop conditions to 
 8-18 

give reasonable assurance that field performance will meet laboratory, testbased expectations.

 • An interpretation and discussion of the test program results by a certified 

coatings engineer. This evaluation should examine, at a minimum, the coating 

performance against the specific tests and the overall requirements for coating 

performance. The overall program must be assessed as to whether it is likely to 

be an effective predictor of actual performance. A recommendation for the use of 

the coating, with specific restrictions, if any, must be included. 

The application should also include general requirements applying to all tests: 

 • Test durations for immersion must equal or exceed the combined maximum 

design (or technical specification) durations for loading and vacuum drying. 

 • An evaluation of any observed gasses, bubbles or other evidence that a gas was 

produced during the test. Coatings that produce flammable gas require a 

mitigation program to prevent burnable or explosive gas concentrations during all 

phases of cask operations. 

8.4.12 Neutron Shielding (MEDIUM Priority) 

8.4.12.1 Neutron Shielding Materials 

Concrete, steel, uranium, and lead typically serve as gamma shields. Boron-filled polymers are 

sometimes used for neutron shielding materials (as opposed to neutron poisons used to control 

criticality). Although dose limits are calculated at the site boundary, not the canister surface, 

these materials are considered ITS, in order to meet the regulatory requirements of 

72.126(a)(6). NUREG/CR-6407 specifically designates neutron shielding materials as ITS 

Category B. 

References for all materials used, including nonstandard materials (e.g., proprietary neutron 

shield material), should be provided for the source of the material composition and density data 

along with validation of the data. The SAR should also describe the geometry of the shielding 

materials. 

In-service performance monitoring of these materials is performed during the required periodic 

radiation surveys. Should a decline in the shielding effectiveness be detected, there is ample 

time and opportunity for engineering evaluation and corrective action. Therefore, the 

qualification and acceptance testing of neutron shielding materials should not be required in the 

TS. 

The SAR should describe the composition of shielding materials and geometries. References 

for all materials used, including nonstandard materials (e.g., proprietary neutron shield material), 

should be provided for the source of the material composition and density data along with 

validation of the data. 

8.4.12.2 Assessing Previously Unreviewed (New) Neutron Shielding Materials 

Should a new material be introduced, review may proceed as follows: 
 8-19 

The reviewer should confirm that temperature-sensitive (e.g., polymeric) neutron shielding 

materials will not be subject to temperatures at or above their design limits during normal 

conditions. The reviewer should determine whether the applicant properly examined the 

potential for shielding material to experience changes in material densities at temperature 

extremes. For example, elevated temperatures may reduce hydrogen content through loss of 

water in concrete or other hydrogenous shielding materials. 

With respect to polymeric neutron shields, the reviewer should verify that the application: 

 • Describes the test(s) demonstrating the neutron-absorbing ability of the shield 

material. 

 • Describes the testing program and provides data and evaluations that 

demonstrate the thermal stability of the resin over its design life while at the 

upper end of the design temperature range. 

• Describes the nature of any temperature-induced degradation and its effect(s) on 

neutron shield performance. 

 • Describes what provisions exist in the neutron shield design to assure that 

excessive neutron streaming will not occur as a result of shrinkage under 

conditions of extreme cold. This description is required because polymers 

generally have a relatively large coefficient of thermal expansion when compared 

to metals. 

 • Describes any changes or substitutions made to the shield material formulation. 

For such changes, describes how they were tested and how that data correlated 

with the original test data regarding neutron absorption, thermal stability, and 

handling properties during mixing and pouring or casting. 

 • Describes the acceptance tests conducted to verify any filled channels used on 

production casks did not have significant voids or defects that could lead to 

greater than calculated dose rates. 

* Describe the materials ability to withstand the combined aging effects of heat and 

radiation field. 

The potential for shielding material to experience changes in material properties at temperature 

extremes should be described in the SAR. Temperature sensitivities of shielding materials 

should be referenced. The SAR should also address degradation from aging, accumulated 

radiation exposure, and manufacturing tolerances. Twenty years of operational experience has 

not resulted in any noticeable decline in the performance of previously accepted materials, as 

verified by examination of periodic radiation survey results on the ISFSI pads at Surry and 

Robinson sites. 

8.4.13 Criticality Control (HIGH Priority) 

U.S. Nuclear Regulatory Commission (NRC) staff reviewer should read 72.104(a), 72.106(b), 

72.124, and 72.236(g). 
 8-20 

Qualification testing is conducted to ensure that (1) the material used will have sufficient 

durability for the application for which it has been designed, (2) the physical characteristics of 

the components of the absorber materials will meet the design requirements, and (3) the 

uniformity of the distribution of 10B is sufficient to meet the requirements of the applications for 

which the absorber materials will be used. Materials that have passed the qualification tests 

must be acceptance tested (See Chapter 10 of this SRP) for use in systems to be used in 

storage or transportation of nuclear fuel. 

8.4.13.1 Neutron-Absorbing/Poison Materials 

Various boron containing materials are used in the nuclear industry as neutron absorbers. 

Since these materials are used in storage containers for fissile materials, the materials should 

have excellent physical and chemical stability, including a high resistance to radiation and 

corrosion. Further, these materials should experience no reduction in effectiveness under 

normal/off-normal and accident-level conditions of storage. Neutron absorbers can consist of 

alloys of boron compounds with aluminum or steel in the form of sheets, plates, rods, liners, and 

pellets. Likewise, neutron absorbers can consist of a core containing mixed aluminum and 

boron carbide (B4C) particles, clad on both sides with aluminum (a composite). 

The neutron absorber material must be demonstrated to be adequately durable for the service 

conditions of the application. These assurances are usually obtained during qualification testing 

of the material. In addition, acceptance tests (see Chapter 10 of this SRP) are performed on 

samples from each production run of the material. This procedure will ensure the properties for 

the plates or other shapes produced are in compliance with the specifications and requirements 

of the application. The uniformity of the distribution of 10B may be addressed in both the 

qualification and the acceptance tests. 

For all boron-containing absorber materials, the reviewer should verify the SAR, with its 

supporting documentation, describes the absorber material's chemical composition, physical 

and mechanical properties, fabrication process, and minimum poison content. The 

manufacturer's data sheet should be submitted to supplement the above information. In the 

case of absorber plates or sheets, the minimum poison content should be specified as an areal 

density (e.g., milligrams of 10B per cm2

). 

For all boron-containing absorber materials, the reviewer should verify that the SAR, with its 

supporting documentation, describes the absorber material's chemical composition, physical 

and mechanical properties, fabrication process, and minimum poison content. If the applicant 

intends to uses an absorber material with a specific trade name, the manufacturer's data sheet 

should be submitted to supplement the above information. In the case of absorber plates or 

sheets, the minimum poison content should be specified as an areal density (e.g., milligrams of 

10B per cm2

). 

8.4.13.2 Computation of Percent Credit for Boron-Based Neutron Absorbers 

This section illustrates one method used by the materials reviewers to compute the level of 

credit to be allowed for 1/v neutron absorber materials, such as boron or lithium, in the criticality 

safety analysis of packages for storing fissile materials, including fresh and SNF. The 

computation of the allowed level of credit uses the results of neutron attenuation measurements 

performed on samples of the absorber material placed in a beam of thermal neutrons. 
 8-21 

Where such validation uncertainties exist, an upper limit of 90 percent credit is applied to boronbased solid absorbers, meaning that the material is computationally modeled as containing only 

90 percent of the 10B shown to be present. The staff has concluded that limiting the poison 

credit to 90 percent adequately accounts for the uncertainties arising in extrapolating the 

validation for boron-based absorber materials. Other remedies, beyond the scope of this 

guidance, may be necessary in addressing the potentially more complex neutron-spectral 

effects and validation uncertainties encountered with materials based on non-1/v-absorbers 

such as cadmium or gadolinium. The current guidance applies only to 1/v absorbers such as 

boron or lithium. 

Neutron channeling has been shown to occur in a commercial product that uses coarse 

particles of natural B4C dispersed in an aluminum matrix. For one material, neutron channeling 

effects reduced the measured attenuation of thermal neutrons by about 18 percent. Therefore, 

whenever uncertainty due to these materials factors exists in a product, it may be necessary to 

measure the neutron attenuation for that product to assess the expected material performance 

in service. Thus, in addition to the 90-percent limit on poison credit that is used to offset 

validation uncertainties, an additional penalty must be considered for material heterogeneity 

effects and uncertainties. In the absence of a fully documented understanding of nonuniformities and channeling effects in a heterogeneous absorber material, the staff recommends 

that the poison credit should continue to be limited to 75 percent.

A neutron absorber material is formulated to meet or exceed the neutron absorption effect 

computed to be required for a given service application. This guidance can be used to extend 

the range of credit for heterogeneous absorber materials from 75 to 90 percent, as follows: 

 • Material for which data is presented to show the measured attenuation for 

thermal neutrons to be at or above the acceptance attenuation (Aa), is given the 

full credit of 90 percent. 

 • Material for which data is presented to show the measured attenuation for 

thermal neutrons to be at levels between 75 and 100 percent of the acceptance 

attenuation (Aa) is given a fraction of the 90 percent credit allowed for fully 

effective absorber material. 

 • Material for which data is presented to show the measured attenuation for 

thermal neutrons to be at or below 75 percent of the acceptance attenuation (Aa) 

is not approved for use at any level of credit; the process used to produce such 

material is judged to be unsuitable. 

The sampling, testing, and reporting of results shall be conducted according to the 

specifications given in ASTM standard C1671-7. 

The applicable credit can be calculated by the following method. Using the following definitions: 

A = neutron attenuation, a measured value taken on a given absorber material in a 

beam of thermal neutrons with fixed energy spectrum. A is assumed to be normally 

distributed with mean : and standard deviation F. 
 8-22 

Aa =Aa = acceptance value of neutron attenuation, based on a qualified homogeneous 

absorber standard such as ZrB2, or a heterogeneous calibration standard that is 

traceable to nationally recognized standards, or calibrated with a monoenergetic 

neutron beam to the known cross section of boron-10. Calibration standards should 

be evaluated at 111 percent (i.e., 1/0.90) of the poison density assumed in the 

criticality computational model. 

At l = attenuation tolerance limit, a statistic of the data 

n = number of coupon measures of attenuation 

P = probability 

μ = true mean of A 

x bar = estimate of μ

F = true standard deviation of A 

S = estimate of F 

Cp = exact number of standard deviations required at probability P 

Kp = tolerance coefficient that is substituted for Cp when : and F are estimated by x bar 

and S, respectively 

( = confidence level 

The attenuation data can be used to bound the probability P that the value of neutron 

attenuation A at an arbitrary location on the material is greater than the acceptance attenuation 

Aa. This is done by computing an attenuation tolerance limit, At l , such that, with 95-percent 

confidence, the probability is less than 0.05 that A < At l. 

Let P = 0.95 and γ = 0.95. Compute At l = (X bar - Kp S), where Kp = f(P, n, γ). The value of KP 

may be found in a table of one-sided tolerance coefficients for a normal distribution.

If At l ≥ Aa, then 90 percent credit is given. 

If At l < Aa, then compute the fractional credit from 0.75 to 0.90 as follows: 

 Fractional Credit = 0.30 + 0.6(At l / Aa). 

If the computed fractional credit is less than 0.75, the process is regarded as unsuitable and 

should be given no credit. 

8.4.13.3 Qualifying the Neutron Absorber Material Fabrication Process 

Not including neutron attenuation, in past reviews the staff has accepted the following 

qualification testing: 

1) Mechanical testing to ensure that the neutron poison material is structurally sound, even 

if the absorber is not used for structural purposes. 

In the past, the staff has accepted ASTM B 557 – 06 tensile testing of samples which 

demonstrated: 

• 0.2% offset yield strength no less than 1.5 ksi 

• ultimate strength no less than 5.0 ksi 

• elongation no less than 1% 
 8-23 

Alternatively, the staff has accepted ASTM E 290 – 97a bend tests, with a 90o

 bend 

without failure as the passing criteria. 

2) Porosity measurements to ensure that the corrosion resistance (which is directly linked 

to hydrogen generation in the spent fuel pool) of the neutron poison material is 

maintained, and that the general structural characteristics of the material are controlled. 

The methodology for porosity is up to the discretion of the applicant. Limits on both the 

total porosity of the material, and the “open” or “interconnected” porosity of the material 

should be explicitly stated in the Technical Specifications. Excluding BoralTM, the total 

open porosity of the neutron poison material should be limited to 0.5 volume percent or 

less. 

3) In general the conditions of spent fuel loading, unloading, and storage do not require 

qualification testing to demonstrate resistance to thermal, radiation, or corrosion induced 

degradation if the neutron absorber is only made of boron carbide and an aluminum 

alloy meeting ASTM chemical requirements for the 1000 or 6000 series of aluminum. 

Other aluminum alloys (particularly those which are not heat-treatable) may also be 

acceptable to the staff without qualification testing. Porosity measurements on the 

neutron poison material should not be waived, regardless of the aluminum alloy used in 

the neutron absorber, however. 

4) A sufficient number of samples should be used to measure the thermal conductivity of 

the neutron poison material at room and elevated temperature. Reviewers should be 

aware that clad neutron poison materials are thermally anisotropic. 

5) For clad materials, a test demonstrating resistance to blistering during the drying 

process should be included in the qualifying tests. In the past the staff has accepted 

testing where: 

Samples of clad materials are soaked in either pure or borated water for 24 hours and 

then insertion into a preheated oven at approximately 825°F for a minimum of 24 hours. 

The samples are then visually inspected for blistering and delamination before 

undergoing qualifying mechanical testing. 

Significant, additional qualifying tests should be conducted for structural neutron poisons. 

Mechanical and thermal tests should include, tensile testing, impact testing (or KIC

measurements), creep testing, and (if applicable) mechanical testing of weldments. 

Samples of neutron poison material should also be examined [i.e., the use of transmission 

electron microscopy (TEM) or scanning electron microscopy (SEM)] for the following changes: 

 • Redistribution or loss of boron. 

 • Dimensional changes (material instability). 

 • Cracking, spalling, or debonding of the matrix from the boron-containing 

particles. 
 8-24 

 • Weight changes caused by leaching, dissolution, corrosion, wear, or off-gassing. 

 • Embrittlement. 

 • Chemical changes such as oxidation or hydriding. 

 • Molecular decomposition of the material as a result of radiation (radiolysis). 

Coupons should be taken so as to be representative of the neutron poison material. To the 

extent practical, test locations on coupons should be stratified to minimize errors due to location 

or position within the coupon. Some suggested locations should include the ends, corners, 

centers, and irregular locations. These locations represent the most likely areas to contain 

variances in thickness. Adequate numbers of samples should be taken from components (i.e., 

plate, rod, etc.) produced from a lot to obtain a good representation. A lot is defined as all plates 

from a single billet. Overall, the coupons should be a representative sample of the material. 

For containers that will be loaded or unloaded in a SNF pool or similar environment, the 

reviewer should verify the absorber material has been evaluated or tested for environmental and 

galvanic interactions and the generation of hydrogen in the pool environment. If environmental 

testing is employed, the test conditions (time, temperature) should equal or exceed those 

expected for loading, unloading, and transfer operations. For environmental tests, the absorber 

materials should be coupled to dissimilar metals, as may be appropriate to the application. The 

environment may be borated or deionized water, as appropriate. The evaluation should also 

consider the effects of any residual pool water remaining in the container after removal from the 

pool. 

Generally, for common engineering materials, an evaluation based upon consultation of a 

corrosion reference (galvanic series) should suffice for pool loading/unloading situations. 

The reviewer should note the applicant must take appropriate measures to assess the strength 

or ductility of the material, depending on the structural requirements of the application. 

Acceptance testing of the fabricated materials is discussed in Chapter 10, “Acceptance Tests 

and Maintenance Program Evaluation,” of this SRP. 

8.4.14 Concrete and Reinforcing Steel (LOW Priority) 

8.4.14.1 Embedment Materials 

The materials discipline should review the material to be used for embedments, inserts, 

conduits, pipes, or other items embedded in the concrete. Embedments must satisfy the 

requirements of the code used in designing the reinforced concrete structure in which they are 

embedded (e.g., ACI 359, ACI 349, or ACI 318). Zinc, zinc rich coatings, zinc-clad materials, 

and aluminum should not be used for any embedded objects in structures designed to ACI 349 

or ACI 359 that will be in contact with wet concrete, because of the potential for concrete 

degradation from an adverse chemical reaction. Embedments and attachments are considered 

to include components cast or grouted into the reinforced concrete structure, inserts, embedded 

pipes, conduits, or lightning protection and grounding systems. 

Unless otherwise specified in this SRP, steel structural attachments must comply with the 

appropriate requirements of ACI-349. 
 8-25 

8.4.14.2 Concrete Temperature Limits 

The NRC accepts the use of ACI 318 for the design and material specifications for reinforced 

concrete structures subject to NRC approval, but they are not important to safety. If ACI 349 is 

used for design of such structures, the NRC accepts the use of ACI 318 for construction. The 

NRC also accepts the following criteria as an alternative to the temperature requirements of ACI 

349 Section A.4, but only for the specified use and temperature ranges: 

 1. Concrete temperatures in general or local areas are a maximum of 93°C (200°F) 

in normal or off-normal conditions and/or occurrences, no tests are needed to 

prove capability for elevated temperatures or reduced concrete strength. 

 2. If concrete temperatures in general or local areas exceed 93°C (200°F) but are 

less than 149°C (300°F), no tests are required to prove capability for elevated 

temperatures or reduced concrete strength if Type II cement is used and 

temperature appropriate aggregates are used. The following criteria for fine and 

coarse aggregates are acceptable: 

- Satisfy ASTM C33 requirements and requirements references in ACI 

 349 for aggregates, and 

- Have a demonstrated coefficient of thermal expansion (tangent in 

 temperature range of 20-38°C (70-100°F) no greater than 11x10-6

 mm/mm/°C (6x10-6 in./in./°F), or be one of the following materials: 

 limestone, dolomite, marble, basalt, granite, gabbro, or rhyolite. 

 • If concrete temperatures in general or local areas under normal or off-normal 

conditions do not exceed 107°C (225°F), the requirements of 1 and 2 (above) 

apply to the coarse aggregate. Fine aggregate that meets 1 (above) and is also 

composed of quartz sands or sandstone sands may be used in place of 2 

(above) and be in compliance. 

8.4.14.3 Omission of Reinforcement 

Frequently, designers specify the omission of reinforcing steel (“rebar”) in concrete aboveground structures which have the purpose of gamma shielding only. This is acceptable since it 

is to avoid the inadvertent formation of voids in the concrete due to the presence of the rebar, 

which can act to block the aggregate in the concrete from filling all intended areas. 

Concrete applied around buried steel structures should be reinforced to alleviate shrinkage 

crack propagation. Concrete alleviates soil corrosion by creating a beneficial chemical buffering 

effect (high pH) around the steel. Cracks allow groundwater plus electrolyte intrusion which 

reduces the effectiveness of the concrete protective barrier. 

8.4.15 Seals 

Applicants for spent fuel storage canisters with metallic seals generally rely on seal 

manufacturer’s data to determine the maximum service temperatures for seals. Seals that may 

potentially be exposed to high temperature may not have been tested by independent 

laboratories (such as NIST and Factory Mutual). Due to the importance of the integrity of the 
 8-26 

seals, laboratory test results or data sheets that reference independent test results should be 

included in applications, if available. 

8.4.15.1 Metallic Seals (MEDIUM Priority) 

Bolted lid canisters employ redundant metallic seals as part of the confinement boundary. 

These seals are ITS components. The primary materials issue is the temperature resistance of 

the seal spring material. Generally this is a nickel-base alloy with excellent temperature and 

creep resistance. The seal cover material may be soft aluminum or silver. Aluminum faced 

seals have failed in service due to corrosion from inadvertent rainwater intrusion. Substitution of 

silver alloy faced seals appears to have alleviated the susceptibility of mechanical seals to this 

corrosion-induced failure mechanism. 

8.4.15.2 Elastomeric Seals (LOW Priority) 

Bolted lid canister designs may also employ a weather cover to preclude rainwater from the 

confinement boundary seals. These weather covers may be sealed against the weather with an 

elastomeric seal such as Viton. As such, these seals may be susceptible to thermally and 

radiation induced aging (hardening). Consequently, a replacement program may be warranted 

if the heat or radiation exposure is sufficient. Guidance as to radiation or thermal resistance is 

usually obtainable from the seal manufacturer. Elastomeric seals have never been ITS 

components in storage canisters. 

Radiation will generally cause polymerization of elastomers to an extent that would adversely 

affect the performance when the dose reaches 105

 Gy (107 rads). For higher dose rate 

environments, elastomer O-rings should not be specified. The use of fluorcarbons, which are 

known to be particularly susceptible to radiation damage, should be restricted if the expected 

dose exceeds 100 Gy (104

 rads). 

The reviewer should verify O-ring seals do not reach their maximum operating temperature limit 

during normal and off-normal conditions of storage. The O-ring manufacturer’s data sheets 

specifying temperature and radiation tolerances should be included in the SAR. 

The materials discipline should review the applicant’s evaluation demonstrating the minimum 

normal operating temperature (usually -40°F) will neither fail the O-ring seal by brittle fracture 

nor stiffen the O-ring (loose elasticity) to an extent that prevents the seal from meeting its 

service requirements. 

The reviewer should verify that under the environmental conditions expected in storage service, 

O-ring seals will not chemically react or decompose in a manner that would significantly affect 

other components of the DSS. 

8.4.16 Low Temperature Ductility and Fracture Control of Ferritic Steels 

(MEDIUM Priority) 

Regulatory Guides 7.11 and 7.12 specify acceptable ferritic steels for low temperature service 

where good toughness is required. Metals having a face-centered cubic crystal structure such 

as austenitic stainless steels remain tough and ductile to very low temperatures and are not a 

concern in this regard. Toughness testing (e.g., Charpy impact) of welds is governed by ASME 

Section III, as supported by Section IX. 
 8-27 

For designs that specify ferritic steels other than those listed in Reg. Guides 7.11 and 7.12, the 

Reg. Guide specifies the types of tests and data needed to qualify a material. Those tests and 

data include dynamic fracture toughness and nil-ductility or fracture appearance transition 

temperature test data. Toughness testing (e.g., Charpy impact) of welds is governed by ASME 

Section III, as supported by Section IX. 

 .

8.4.17 Cladding 

(MEDIUM Priority) This guidance will allow all commercial spent fuel that is currently licensed by 

the Nuclear Regulatory Commission (NRC) for commercial power plant operations to be stored 

in accordance with the regulations contained in 10 CFR Part 72. However, cask vendors' 

requests for the storage of spent fuel with burnup levels in excess of those levels licensed by 

the Office of Nuclear Reactor Regulation (NRR), or for cladding materials not licensed by NRR, 

may require additional justifications by the applicant. 

The most important issues regarding spent fuel and cladding that must be considered are: 

 • The maximum cladding temperature during loading/unloading operations and 

normal conditions of storage. For high burn-up fuel, defined as any fuel with a 

burn-up greater than 45GWd/MTU, the maximum allowable cladding temperature 

limit is 400°C. For materials analyses, an appropriate maximum fuel burn-up is 

to be specified as the peak rod average. 

 • Compatibility of fuel bundle materials and non-fuel component materials such as 

burnable poison rod assemblies (BPRAs) with the loading/unloading environment 

and the cask interior components. Refer to the separate discussion of this in 

Section 8.4.8.1. 

 • The fuel is maintained in a water or inert environment during loading/unloading 

operations to prevent excessive oxidation of fuel pellets. This is discussed in 

more detail in Section 8.7 of this SRP. 

 • A definition of damaged fuel is adequate for the intended fuel load and fuel with 

more severe damage (if any) is precluded from loading. 

8.4.17.1 Cladding Temperature Limits (MEDIUM Priority) 

The requirements of 10 CFR 72.122(h)(1) seek to ensure safe fuel storage and handling and to 

minimize post-operational safety problems with respect to the removal of the fuel from storage. 

In accordance with this regulation, the spent fuel cladding must be protected during storage 

against degradation that leads to gross rupture of the fuel and must be otherwise confined such 

that degradation of the fuel during storage will not pose operational problems with respect to its 

removal from storage. Additionally, 10 CFR 72.122(l) and 72.236(m) require that the storage 

system be designed to allow ready retrieval of the spent fuel from the storage system for further 

processing or disposal. 

Spent fuel storage casks and systems must be designed to meet four safety objectives: 

 • Ensure doses from the spent fuel in the casks and systems are less than limits 

prescribed in the regulations. 
 8-28 

 • Maintain subcriticality under all credible conditions. 

 • Ensure there is adequate confinement and containment of the spent fuel under 

all credible conditions of storage. 

 • Allow the ready retrieval of the spent fuel from the storage systems. 

The acceptance criteria below and review procedures are designed to provide reasonable 

assurance the spent fuel is maintained in the configuration analyzed in the storage SARs. 

These criteria are applicable to all commercial spent fuel burnup levels and cladding materials. 

In order to assure integrity of the cladding material, the following criteria should be met: 

 • For all fuel burnups (low and high), the maximum calculated fuel cladding 

temperature should not exceed 400°C (752°F) for normal conditions of storage 

and short-term loading operations (e.g., drying, backfilling with inert gas, and 

transfer of the cask to the storage pad). However, for low burnup fuel, a higher 

short-term temperature limit may be used, if the applicant can show by 

calculation the best estimate cladding hoop stress is equal to or less than 

90 MPa (13,053 psi) for the temperature limit proposed. 

• During loading operations, repeated thermal cycling (repeated heatup/cooldown 

cycles) may occur but should be limited to less than 10 cycles, where cladding 

temperature variations are more than 65°C (117°F) each. 

• For off-normal and accident conditions, the maximum cladding temperature 

should not exceed 570°C (1058°F). 

Given the conservatism used in calculating peak clad temperatures for low burnup fuel, the staff 

has reasonable assurance that storage cask systems which use the 570°C temperature limit for 

low burnup fuel loading operations will continue to perform as expected when the casks were 

originally certified. Therefore, there is no need to require the licensees of storage-only or dualpurpose cask systems to repackage spent fuel loaded using the 570°C temperature limit. 

The maximum allowable temperature should be based upon the peak rod temperature, not the 

average rod temperature. By employing the peak rod temperature, only a small fraction of the 

rods will experience the temperature and stress conditions that could lead to the formation of 

radial hydrides during normal conditions of storage. 

High burnup fuel (i.e., fuel with burnups generally exceeding 45 GWd/MTU) may have cladding 

walls that have become relatively thin from in-reactor formation of oxides or zirconium hydride. 

For design basis accidents, where the structural integrity of the cladding is evaluated, the 

applicant should specify the maximum cladding oxide thickness and the expected thickness of 

the hydride layer (or rim). Cladding stress calculations should use an effective cladding 

thickness that is reduced by those amounts. The reviewer should verify that the applicant has 

used a value of cladding oxide thickness that is justified by the use of oxide thickness 

measurements, computer codes validated using experimentally measured oxide thickness data, 

or other means that the staff finds appropriate. Note that oxidation may not be of a uniform 

thickness along the axial length of the fuel rods. 
 8-29 

Since the hoop stress is dependent on the rod internal pressure, cladding geometry, and the 

temperature of the gases inside the rod, the staff will verify that the applicant has calculated the 

best estimate hoop stress corresponding to the rod internal pressure of the highest burnup fuel 

assemblies of the specific type of assembly. 

The intent of the thermal cycling acceptance criteria is to prevent licensees from applying cask 

drying, loading and transfer operations that could inadvertently enhance an undesirable hydride 

reorientation to form radial hydrides. Accordingly, these criteria pertain only to periods of fuel 

loading and transfer operations of the casks to the storage pads. 

In general, the materials reviewer should coordinate with the structural reviewer to assure the 

spent fuel is maintained in the configuration analyzed in the Safety Analysis Reports (SARs) in 

order to meet the objectives described above. 

The materials reviewer should coordinate with the thermal reviewer to assure the temperature 

criteria stated above are met. If higher peak temperatures are proposed by the applicant, 

additional justification for the higher temperatures must be supplied. 

This guidance will allow all commercial spent fuel that is currently licensed by the Nuclear 

Regulatory Commission (NRC) for commercial power plant operations to be stored in 

accordance with the regulations contained in 10 CFR Part 72. However, cask vendors' requests 

for the storage of spent fuel with burnup levels in excess of those levels licensed by the Office of 

Nuclear Reactor Regulation (NRR), or for cladding materials not licensed by NRR, may require 

additional justifications by the applicant. 

Background justification for these temperature limits can be found in Sec 8.8 of this SRP. 

8.4.17.2 Fuel Classification (HIGH Priority) 

The staff should verify that the definitions below are used in the SAR, and where appropriate 

are also included in the CoC. 

Spent Nuclear Fuel (SNF) - See 10 CFR Part 72.3 for definition. This term has been used in the 

nuclear industry, at different times, to mean the fuel pellets, the rod, or entire fuel assembly. 

Unless specifically modified, the term will refer to both the rods and fuel assembly. 

Damaged SNF - Any fuel rod or fuel assembly that cannot fulfill its fuel-specific or systemrelated functions. 

Undamaged SNF - SNF that can meet all fuel-specific and system-related functions. As 

shown in Figure 8-2, undamaged fuel may be breached. Fuel assembly classified as 

undamaged SNF may have “assembly defects.” 

Breached spent fuel rod - Spent fuel rod with cladding defects that permit the release of gas 

from the interior of the fuel rod. A breached spent fuel rod may also have cladding defects 

sufficient to permit the release of fuel particulate. A breach may be limited to a pinhole leak, 

hairline crack, or may be a gross breach. 

Pinhole leaks or hairline cracks - Minor cladding defects that will not permit significant release of 

particulate matter from the spent fuel rod, and therefore present a minimal as low-as-is-
 8-30 

reasonably-achievable concern, during fuel handling operations. (See discussion of gross 

defects for size concerns.) 

Grossly breached spent fuel rod - A subset of breached rods. A breach in spent fuel cladding 

that is larger than a pinhole leak or a hairline crack. An acceptable examination for a gross 

breach is a visual examination that has the capability to determine the fuel pellet surface may be 

seen through the breached portion of the cladding. Alternatively, review of reactor operating 

records may provide evidence of the presence of heavy metal isotopes indicating that a fuel rod 

is grossly breached. (See discussion for size concerns.) 

Intact SNF - Any fuel that can fulfill all fuel-specific and system-related functions and that is not 

breached. Note that all intact SNF is undamaged, but not all undamaged fuel is intact, since 

under most situations, breached spent fuel rods that are not grossly breached will be considered 

undamaged. 

Can for Damaged Fuel - A metal enclosure that is sized to confine one damaged spent fuel 

assembly. A fuel can for damaged spent fuel with damaged spent-fuel assembly contents must 

satisfy fuel-specific and system-related functions for undamaged SNF required by the applicable 

regulations. 

Assembly Defect - Any change in the physical as-built condition of the assembly with the 

exception of normal in-reactor changes such as elongation from irradiation growth or assembly 

bow. Examples of assembly defects: (a) missing rods; (b) broken or missing grids or grid 

straps (spacers); and (c) missing or broken grid springs, etc. An assembly with a defect is 

damaged only if it can't meet its fuel-specific and system-related functions required by the 

applicable regulations. 

A fuel-specific regulation - a characteristic or performance requirement of the fuel specifically 

named in the applicable Code of Federal Regulations (CFR). These are regulations that specify 

capabilities that the spent nuclear fuel (SNF) must have. Examples include 10 CFR 

72.122(h)(1) and 10 CFR 72.122(l). 

A system-related regulation - a performance requirement placed on the fuel so that the storage 

system can meet its regulatory requirements. Examples include 10 CFR 72.122(h)(5) and 

10 CFR 72.124(a).. 

Previous definitions of damaged fuel have identified specific characteristics of the fuel that 

classify it as damaged, irrespective of whether the fuel is being stored or transported and 

independent of the design of the storage or transportation system. The current staff position is 

that damaged fuel is defined in terms of the characteristics needed to perform the fuel-specific 

and system-related functions. The materials properties, and possibly the physical condition, of 

a fuel rod or assembly can be altered during irradiation or storage. If this alteration is large 

enough to prevent the fuel or assembly from performing its fuel-specific or system-related 

functions during storage, then the fuel assembly is considered damaged. 

To determine whether a fuel assembly is undamaged, the following should be stated in the 

SAR: 

1) The functions the applicant has imposed on the fuel rods and assembly by either fuel 

specific or system-related functions to meet a regulatory requirement for the designated 

phase (storage, transportation, or both); 
 8-31 

2) The mechanisms of change (alteration mechanisms) or the characteristics of the fuel 

that could potentially cause the fuel to fail to meet its fuel-specific or system-related 

functions; 

3) An acceptable analysis showing that the fuel with the designated characteristics will 

meet the fuel-specific and system-related functions when the mechanisms considered in 

item #2, above, are evaluated; and 

4) The physical characteristics of the fuel, based on item #3, above, that could cause the 

fuel or assembly to be classified as "damaged." 

A "default" definition of damaged SNF, derived from ANSI N14.33-2005, is provided for those 

that do not want to perform the assessment outlined in item numbers 1 through 4 above. The 

default definition, however, may not take full advantage of the flexibility of the performancebased definition of damaged fuel provided in this guidance. This default definition may be more 

restrictive than necessary, depending on the design of the storage or transportation cask. For 

example, the default definition of damaged SNF indicates that SNF must be classified as 

damaged if an individual fuel rod is missing from an assembly. However, if an analysis shows 

that all fuel-specific and system-related functions will be met (e.g., subcriticality will be 

maintained, that the SNF assembly will be retrievable and that the structural properties of the 

assembly are not compromised by the missing rod) the assembly may be classified as 

undamaged. An alternative default definition of damaged Spent Nuclear Fuel (SNF) is: SNF 

assemblies must be classified as damaged if any one of the following conditions exist: 

On removal of SNF selected for dry storage or transport from the spent fuel pool, any of the 

following apply: 

 • There is visible deformation of the rods in the SNF assembly. Note: This is not 

referring to the uniform bowing that occurs in the reactor. This refers to bowing 

that significantly opens up the lattice spacing. 

 • Individual fuel rods are missing from the assembly. Note: The assembly may be 

reclassified as intact if a dummy rod that displaces a volume equal to, or greater 

than, the original fuel rod, is placed in the empty rod location. 

 • The SNF assembly has missing, displaced, or damaged structural components 

such that either: 

a. Radiological and/or criticality safety is adversely affected (e.g., 

significantly changed rod pitch). 

b. The assembly cannot be handled by normal means (i.e., crane and 

grapple). 

 • Reactor operating records (or other records) indicate that the SNF assembly 

contains fuel rods with gross breaches. 

 • The SNF assembly is no longer in the form of an intact fuel bundle (e.g., consists 

of, or contains, debris such, as loose fuel pellets or rod segments). 
 8-32 

Additional background and examples of defining damaged fuel can be found in Section 8.6 of 

this SRP. 

8.4.17.3 Reflood Analysis (HIGH Priority) 

The NRC accepts that the total stress on the cladding is maintained below the material’s 

minimum yield stress. The total stress includes the thermal stress combined with the cladding 

hoop stress from internal rod pressure and the rod-gas plenum temperature. The analysis also 

should account for high burnup effects on the fuel (e.g., waterside corrosion, high internal rod 

pressure) and minimum manufacturing wall thickness. Other assembly components should also 

be examined in a similar manner. Engineering judgment, combined with relevant industry 

operational experience with unloading SNF from transportation and storage casks, may support 

the basis for limits on quench fluid temperature and flow rate. This review should be 

coordinated with the thermal reviewer. 

8.4.18 Prevention of Oxidation Damage During Loading of Fuel (MEDIUM Priority) 

The guidance in this section is only applicable to irradiated LWR fuel or other uranium oxide 

based fuel. The reviewer should make sure that the oxidation of other types of fuels during 

loading is evaluated. The information given in this section and Section 8.7 of this SRP may not 

be applicable to other fuel types. The characteristics of those fuel types must be considered 

when evaluating their analysis. 

Once the fuel rods are placed inside of the storage cask and water is removed to a level that 

exposes any part of the rods to a gaseous atmosphere, reasonable assurance the spent fuel 

cladding will be protected against splitting due to fuel oxidation might occur must be 

demonstrated. If oxidation occurred, it may lead to loss of retrievability, or to a configuration not 

adequately analyzed for radiation dose rates or criticality safety. Further, the release of fuel 

fines or grain-sized powder into the inner cask environment from ruptured fuel may be a 

condition outside the licensing basis for the cask system. Three possible options exist to 

address the potential for and consequences of fuel oxidation: 

1. Maintain the fuel rods in an appropriate environment such as Ar, N2, or He to prevent 

oxidation. 

 

2. Assure there are not any cladding breaches (including hairline cracks and pinhole leaks) 

in the fuel pin sections that will be exposed to an oxidizing atmosphere. This can be 

done by a review of records (for example, sipping records) or 100 percent eddy current 

inspection of assemblies. 

3. Determine the time-at-temperature profile of the rods while they are exposed to an 

oxidizing atmosphere and calculate the expected oxidation to determine if a gross 

breach would occur. The analysis should indicate the time required to incubate the 

splitting process will not be exceeded. Such an analysis would have to address 

expected differences in characteristics between the fuel to be loaded and the fuel tested 

to determine the basis for the analysis. Conversely, the maximum allowable 

temperature of the rods could be limited to the temperature that calculations show 

cladding splitting will not be expected to occur. Such evaluations must incorporate the 

effects of uncertainty in the data base. Calculation of the possibility of cladding splitting, 

is fraught with all the uncertainties discussed above. Lowering the maximum allowable 

temperature may impose an economic penalty by limiting the heat load in the cask. The 
 8-33 

selection of the methodology used to address this issue is up to the applicant. The use 

of a non-oxidizing atmosphere in the fuel canister to prevent fuel oxidation is one method 

accepted by the staff to address the issue. 

If Option 3 is chosen, the materials reviewer should coordinate with the thermal reviewer to 

determine that the operating procedures, technical specification, and associated licensing 

documentation, as submitted by the applicants, provide a supportable analysis of the potential 

for cladding splitting, should fuel rods be exposed to an oxidizing gaseous atmosphere. For fuel 

with burnup below ~45 GWd/MTU and Zircaloy cladding, the time-at-temperature (TT) curves 

developed to date (R.E. Einziger and R.V. Strain, “Oxidation of Spent Fuel at Between 250° and 

360°C,” EPRI Report NP-4524, 1986, for example) can be used to determine the allowable 

exposure duration to an oxidizing atmosphere if the fuel temperature is known, or conversely 

the maximum allowable temperature if the exposure time is known. For example, using 

Figure 3-9 of the above reference, at 360°C one would expect to incur splitting between 2 and 

10 hours. On the other hand, if one expected to stay at temperature for 100 hours then the fuel 

temperature should be kept below 290°C. 

Additional information on oxidation of damaged fuel can be found in Section 8.7 of this SRP. 

Please refer to this reference for additional detail and background. 

 

 8.4.19 Flammable Gas Generation (MEDIUM Priority)

The reviewer should assume the generation of hydrogen or other gases during wet 

loading/unloading operations occurs. Field experience has amply demonstrated that any 

canister design employing aluminum components as part of the fuel basket construction will 

have a propensity to generate hydrogen. Efforts to passivate the aluminum components have 

proven inadequate to eliminate the generation of hydrogen. The use of zinc, zinc-rich coatings, 

or zinc-clad materials (e.g., galvanized steel) in particular, is known to generate potentially large 

quantities of hydrogen gas during wet-loading in SFP. 

 Consequently, the reviewer should verify the operating procedures contain adequate guidance 

for detecting the presence of hydrogen and preventing the ignition of combustible gases during 

cask loading and unloading operations. These procedures must be incorporated by reference 

into the TS. 

 8.4.20 Canister Closure Welds Testing (MEDIUM Priority)

Helium leakage testing of the entire confinement boundary is performed to demonstrate 

compatibility with the design basis leak rate, and ensures that: 

 • the fuel payload is protected from the deleterious oxidizing effects of moisture by 

excluding intrusion of such, 

 • the helium inerting gas will remain in the canister in sufficient amount over the 

license period, and 

 • the helium gas heat transfer medium will remain in sufficient quantity over the 

license period to assure the cladding temperatures are controlled at safe levels. 

This guidance addresses all welds associated with the redundant closures of a spent fuel 

canister and describes how each individual closure weld must be considered from the overall 
 8-34 

design and testing standpoint. It only applies to canisters of all-welded construction, fabricated 

from austenitic stainless steel, employing redundant welds for the confinement closure. 

The staff should verify that the cask design under review is in compliance with the guidance of 

this document. In order for any closure weld to be exempt from the helium leak testing to 

demonstrate compliance with 10 CFR 72.236, the staff should verify all of the following 

conditions are satisfied: 

 • The welded canister (i.e., the confinement boundary) must be leak tested in 

accordance with ANSI N14.5-1997, except as specified by this guidance. 

 • Closure welds must conform with the guidance of this SRP, as appropriate. 

 • “Structures, systems, and components important to safety must be designed to 

withstand postulated accidents.” [10 CFR 72.122(b)(1)]. 

 • Records documenting the lid welds shall comply with the provisions of 10 CFR 

Part 72.174, “Quality Assurance Records.” Records storage should comply with 

ANSI N45.2.9, “Requirements for Collection, Storage, and Maintenance of 

Quality Assurance Records for Nuclear Power Plants.” 

 • Activities related to inspection, evaluation, documentation of fabrication, and lid 

welding shall be performed in accordance with an NRC-approved quality 

assurance program as required in 10 CFR Part 72, Subpart G, “Quality 

Assurance.” 

In addition for exemption of large multi-pass welds from helium leak testing the following must 

be satisfied. 

 (1) The weld must be multi-pass, with a minimum weld depth comprised of at least 3 

distinct weld layers. 

 (2) Each layer of weld may be composed of one or more adjacent weld beads. 

 (3) The layer must be complete across the width of the weld joint. 

 (4) If only 3 weld layers comprise the full thickness of the weld, each layer must be 

PT examined. 

 (5) For more than 3 weld layers, not all weld layers need be PT examined. The 

maximum weld deposit depth allowed before a PT examination is necessary is 

based upon flaw-tolerance calculations in accordance with Section 8.9 of this 

SRP. Note: This criterion does not supersede the flaw acceptance criteria of any 

construction code. Instead, this criterion is used to establish the maximum 

allowable weld deposit depth before an in-process PT examination is necessary. 

 (6) Regardless of conditions (4) or (5) above, at least 3 different weld layers must be 

examined, e.g., the root pass, a mid-layer, and the cover pass. 

 (7) The weld cannot have been executed under conditions where the root pass 

might have been subjected to pressurization from the helium fill in the canister 
 8-35 

itself. Credit may not be taken for closure valves, quick-disconnects, or similar. 

It is assumed that mechanical closure devices (e.g., a valve or quick-disconnect) 

permit helium leaks. Practical experience has shown such leaks occur and have 

been responsible for causing leak paths through the weld. Consequently, welds 

potentially subjected to helium pressure (by way of leakage through a 

mechanical closure device) during the welding process must be subsequently 

helium leak tested. 

Other closure issues the materials reviewer should evaluate are: Hydrostatic Testing, ASME 

Code Case N-595-4, and the limiting root pass criteria for the weld. 

Closure welds must be hydrostatically or pneumatically tested in accordance with ASME Code 

Section III requirements to the extent practicable. The two designs discussed in Section 8.9 of 

this SRP meet this criterion. 

ASME Code Case N-595-4 is not endorsed by the NRC staff, per Regulatory Guide (RG) 1.193 

and consequently is not permitted as an alternative to the Code requirements. 

Cask lid welding is governed in part by the limiting flaw size analysis. The welding method 

described herein controls the depth of weld deposit for the intermediate passes before the 

required PT examination is performed. However, the root pass thickness is not addressed by 

this guidance, as a single layer root pass was assumed. Occasionally, multi-layer root passes 

are employed to smooth the weld surface to avoid false positives from the PT. 

A multi-layer root pass is acceptable provided the same method of limiting the weld deposit 

depth is followed as for the intermediate weld passes. Stress analysts should note that the 

intermediate layer critical flaw size calculation assumes a buried flaw, not a surface connected 

flaw. For the root pass calculation, a surface connected flaw must be assumed. This will result 

in a smaller critical flaw size, and, consequently a smaller permissible weld deposit thickness 

before a PT exam is considered necessary. 

The staff should verify that if the licensee desires to use a thicker root pass, they must limit the 

amount of weld deposit to the ratio of the fracture toughness K values (or, J values) for the 

different flaw types (buried K divided by surface K) multiplied by the maximum depth. This will 

limit the depth of the root pass to the critical flaw size for a surface connected flaw. Thus, if a 

licensee desires to use a thicker weld deposit for the root pass, then a limiting flaw size analysis 

establishes a structural basis. 

The staff recognizes that for stainless steel, K, or even J, is not entirely correct for evaluating 

failure in austenitic stainless steel due to the large capacity for plastic deformation. Generally 

the result is failure due to net section stress, not fracture. However, the stress intensity ratio 

suggested above is acceptable for this purpose. 

The regulatory requirements governing this review are: 10 CFR 72 122(a), 72.122(h)(5), 

72.104(a), 72.106(b), 72.236(d), 72.236(e), 72.236(j), and 72.236(l). 

Please refer to the additional information in Section 8.9 of this SRP to supplement the review 

criteria. 

8.4.21 Periodic Inspections (LOW Priority) 
 8-36 

Review the SAR operations and acceptance testing chapters for appropriate periodic inspection 

programs which may be included for the purpose of monitoring materials conditions or 

performance. Some cask vendors are now anticipating future license renewal for the designs 

and are incorporating into the SAR the currently specified limited inspections that are required 

as part of a license renewal application. 

 • A one-time inspection of normally inaccessible portions of the canister exterior for 

unanticipated corrosion or other degradation. A single canister (or several) may 

be selected based upon engineering criteria such as longest time in service, 

hottest operating temperature, etc. and used to “bound” other canisters of that 

type of material of construction. 

 • The periodic (usually monthly) ISFSI radiation survey results should be reviewed 

to determine if any significant degradation of any neutron shielding material (if 

used) has occurred. 

8.5 Evaluation Findings 

The evaluation findings are prepared by the reviewer on satisfaction of the regulatory 

requirements of Section 8.3. The reviewer should examine these requirements and provide a 

summary statement for each. These statements should be similar to the following examples: 

 F8.1 Section(s) of the SAR adequately describe(s) the materials used for SSCs 

important to safety and the suitability of those materials for their intended 

functions in sufficient detail to evaluate their effectiveness. 

 F8.2 The applicant has met the requirements of 10 CFR 72.122(a). The material 

properties of SSCs important to safety conform to quality standards 

commensurate with their safety function. 

 F8.3 The applicant has met the requirements of 10 CFR 72.104(a), 72.106(b), and 

72.124. Materials used for criticality control and shielding are adequately 

designed and specified to perform their intended function. 

 F8.4 The applicant has met the requirements of 10 CFR 72.122(h)(1) and 72.236(h). 

The design of the DSS and the selection of materials adequately protect the SNF 

cladding against degradation that might otherwise lead to damaged fuel. 

 F8.5 The applicant has met the requirements of 10 CFR 72.236(h) and 72.236(m). 

The material properties of SSCs important to safety will be maintained during 

normal, off-normal, and accident conditions of operation so the SNF can be 

readily retrieved without posing operational safety problems. 

 F8.6 The applicant has met the requirements of 10 CFR 72.236(g). The material 

properties of SSCs important to safety will be maintained during all conditions of 

operation so the SNF can be safely stored for the minimum required years and 

maintenance can be conducted as required. 

 F8.7 The applicant has met the requirements of 10 CFR 72.236(h). The [cask 

designation] employs materials that are compatible with wet and dry SNF loading 
 8-37 

and unloading operations and facilities. These materials should not degrade 

over time or react with one another during any conditions of storage. 

The reviewer should provide a summary statement similar to the following: 

“The staff concludes the material properties of the structures, systems, and components 

of the [cask designation] are in compliance with 10 CFR Part 72, and that the applicable 

design and acceptance criteria have been satisfied. The evaluation of the material 

properties provides reasonable assurance the [cask designation] will allow safe storage 

of SNF for a licensed (certified) life of years. This finding is reached on the basis 

of a review that considered the regulation itself, appropriate regulatory guides, applicable 

codes and standards, and accepted engineering practices.” 

8.6 Supplemental Information for Methods for Classifying Fuel (HIGH Priority) 

A. Grossly Breached SNF Cladding 

The regulations in 10 CFR 72.122(h) state “The spent fuel cladding must be protected during 

storage against degradation that leads to gross ruptures or the fuel must be otherwise confined 

such that degradation of the fuel during storage will not pose operational safety problems with 

respect to its removal from storage.” 

In dry cask storage and transportation systems, a gross cladding breach should be considered 

as any cladding breach that could lead to the release of fuel particulate greater than the average 

size fuel fragment. A pellet is ~1.1 centimeters in diameter in 15 x 15 Pressurized-Water 

Reactor (PWR) assemblies. Pellets from a Boiling-Water Reactor (BWR) are somewhat larger, 

and those from 17 x 17 PWR assemblies are somewhat smaller. The pellet's length is slightly 

longer than its diameter. During the first cycle of irradiation in-reactor, the pellet fragments into 

25-35 smaller interlocked pieces, plus a small amount of finer powder, due to, pellet-to-pellet 

abrasion. When the rod breaches, about 0.1 gram of this fine powder may be carried out of the 

fuel rod at the breach site. Modeling the fragments as either spherical- or pie-shaped pieces 

indicates that a cladding-crack width of at least 2-3 millimeters would be required to release a 

fragment. Hence, gross breaches should be considered to be any cladding breach greater than 

1 millimeter. 

A review of reactor operating records, ultrasonic testing, and sipping (if done in a timely fashion) 

can be used to classify rods as unbreached or, breached. Evidence of only gaseous or volatile 

decay products (no heavy metals) in the reactor coolant system is accepted as evidence that a 

cladding breach is no larger than a pinhole leak or hairline crack. Records that show the 

presence of heavy metal isotopes that are characteristic of fuel release in the reactor coolant 

system indicate gross breaches in the cladding. Likewise, visual examination may also be used 

to determine if a cladding breach is gross, if the breached rod can be positively identified. 

Because cladding openings larger than 1 millimeter should expose the fuel pellet to visual 

sighting, visual examination of the breached rod can be used to determine if a breach is gross. 

However, visual examination is not an acceptable method of confirming intact (undamaged) fuel 

for assemblies that have indicated leakage. 

It should be noted, however, that undamaged spent-fuel rods with pinhole leaks and/or hairline 

cracks will expose the fuel pellets to the canister or cask atmosphere. If that atmosphere is 

oxidizing, then the fuel pellet may oxidize and expand, placing stress on the cladding. The 

expansion may eventually cause a large split in the cladding, resulting in spent fuel that must be 
 8-38 

classified as damaged (for storage and possibly also for transportation) due to gross breaches 

in the cladding. Since fuel oxidation and cladding splitting follow Arrhenius time-at-temperature 

behavior, fuel rods with pinholes or hairline cracks that are exposed to an oxidizing atmosphere 

may experience this type of additional cladding damage. Section 8.7 of this SRP, 

“Supplemental Information for Potential Rod Splitting Due to Exposure to an Oxidizing 

Atmosphere during Short-Term Cask Loading Operations in LWR or other Uranium Oxide 

Based Fuel,” provides information regarding prevention of this phenomenon. Before handling 

undamaged rods with pinhole leaks and/or hairline cracks in an oxidizing atmosphere, the 

potential fuel and cladding degradation at the temperature of interest for the duration of the 

process should be assessed. 

B. Fuel Assembly with Defects 

Damage under this guidance refers to alterations of the fuel assembly that prevent it from 

fulfilling its fuel-specific or system-related functions. Defects such as dents in rods, bent or 

missing structural members, small cracks in structural members, missing rods, etc., need not be 

considered damaged if the applicant can show that the fuel assembly with these defects still 

fulfils its fuel-specific and system-related functions. This may be done using calculations based 

on approved codes, situation-specific data, or reasoned engineering arguments. 

C. Canning Damaged Fuel 

Spent fuel that has been classified as damaged for storage must be placed in a can designed 

for damaged fuel, or in an acceptable alternative. The purpose of a can designed for damaged 

fuel is to (1) confine gross fuel particles, debris, or damaged assemblies to a known volume 

within the cask; (2) to demonstrate that compliance with the criticality, shielding, thermal, and 

structural requirements are met; and (3) permit normal handling and retrieval from the cask. 

The can designed for damaged fuel may need to contain neutron-absorbing materials, if results 

of the criticality safety analysis depend on the neutron absorber to meet the requirements of 

10 CFR 72.124(a). 

D. Relationship of Spent Fuel Populations 

The applicant will designate the population of spent fuel for which the cask system was 

designed (e.g., type of fuel, minimum cooling time, burnup limitations, arrays, manufacturers, 

cladding types, etc.). This population may contain breached rods. Some of these breached 

rods may be grossly breached. It may also contain assemblies with defects, such as missing 

rods, missing grid spacers, or damaged spacers. The populations of breached rods, grossly 

breached rods, and assemblies with defects are determined by in-reactor behavior and exreactor handling. 

Each of these populations must be classified as damaged or undamaged after the storage or 

transportation system has been designated. For example, an applicant might propose the use 

of air as a cover gas in its design of a storage cask. The applicant might also propose this cask 

for use in storing spent fuel with cladding breaches that are hairline cracks or pinhole leaks. 

However, if the spent fuel in the cask will operate at a sufficiently high temperature for a long 

enough time, then oxidation of fuel pellets in breached rods could occur resulting in gross 

breaches. If this is the case, the breached spent fuel should be considered damaged because 

grossly breached rods do not meet the requirements of 10 CFR 72.122(h)(1). If an inert 

atmosphere was used instead of air, only grossly breached rods would be considered damaged 

for storage. This concept is illustrated in Figure 8-2, "Relationship of Spent Fuel Populations." 
 8-39 

Example of Methodology 

The following example is given to illustrate the general methodology. This is only an example of 

the methodology and should not be construed as approved characterization of damaged fuel. 

Figure 8-2 Relationship of Spent Fuel Populations 

Example of Methodology: 

Situation - The vendor of a dual-purpose cask wants to store and transport low-burnup PWR 

fuel in an inert atmosphere and within the temperature limits recommended in Section 8.4.17.1. 

The vendor wants to store assemblies having rods with breaches containing only pinholes or 

hairline cracks, and assemblies having one or more outer grid straps with defects at three or 

more grid locations without canning them. The vendor is only applying for a storage license at 

this time but wants to be reasonably certain that the fuel will also be transportable. 

Activity - Storage of Spent Fuel 

Fuel-specific or system-related functions imposed on rods and assemblies - 10 CFR 

72.122(h)(1), regarding gross ruptures, and 10 CFR 72.122(l), concerning retrievability, must be 

met for storage. 10 CFR 71.55(d), requiring the system to remain subcritical and unchanged 

during normal transport, must be met. The vendor believes that all the remaining system 

requirements, except for the subcriticality requirement, can be met, without imposing any 

limitations on the fuel, if the fuel is within the bounds stated in the situation. 
 8-40 

Mechanisms - There are no mechanisms for the pinhole leaks and hairline cracks to evolve into 

gross breaches since the atmosphere is inert and the temperature is controlled. To be 

retrievable, the assemblies with missing grid straps must be able to withstand design basis 

events in a storage cask. Since the applicant also wants these assemblies to be considered 

undamaged for transportation, the behavior of the assemblies under both normal and 

hypothetical accident transportation conditions in 10 CFR Part 71 must be evaluated. For 

example, for normal transportation conditions, the applicant must show that the assemblies with 

the most missing grid straps in the worst locations can withstand both normal vibration and a 

one-foot drop and remain in their original physical configuration. Additionally, for hypothetical 

accident conditions, the analysis must indicate, among other things, that the system will meet 

shielding and subcriticality requirements when placed under the mechanical and thermal loads 

specified in 10 CFR Part 71. 

Analysis - The applicant conducts an analysis to satisfactorily demonstrate that the assembly 

with three missing grid straps in the worst configuration remains intact for 1) normal 

transportation conditions; 2) cask tip-over; and 3) regulatory accident conditions. Further 

acceptable analysis indicates that all the system-related regulations are met, if the fuel with the 

characteristic limitations (as noted in Characteristics section below), stays structurally intact. 

Characteristics - Assemblies containing breached rods with up to three grid straps missing will 

be considered undamaged for the purposes of storage. Analysis shows that these assemblies 
 8-41 

could probably also be considered undamaged for transportation, but fuel with these 

characteristics will be evaluated and approved as part of a later application for the transportation 

cask certification. 

8.7 Supplemental Information for Potential Rod Splitting Due to Exposure to an 

Oxidizing Atmosphere During Short-Term Cask Loading Operations in LWR or 

Other Uranium Oxide Based Fuel (MEDIUM Priority) 

The definition of undamaged fuel includes fuel rods containing no cladding defects greater than 

pinhole leaks or hairline cracks. During the cask water removal process parts of, or all of, the 

fuel rods will be exposed to a gaseous atmosphere. If the gaseous atmosphere is oxidizing, 

oxidation of fuel pellets or fuel fragments can occur if a cladding breach exists (such as a 

pinhole). Oxidation may occur rapidly and cause significant swelling of fuel pellets and 

fragments, which could result in gross fuel cladding breaches if the time-at-elevatedtemperature after water removal is excessive. 

8.7.1 Fuel Oxidation and Cladding Splitting 

Irradiated uranium dioxide exposed to an oxidizing atmosphere will eventually oxidize to U3O8. 

The time it takes to oxidize is a function of temperature that follows an Arrhenius function and

burnup. However, at temperatures that may be expected for some spent fuel, this reaction can 

occur within a matter of hours. 

 

The grain boundaries of irradiated fuel are highly populated with voids and gas bubbles. Initially 

the grain boundaries are oxidized to U4O9 resulting in a slight matrix shrinkage and further 

opening of the pellet structure. Oxidation then proceeds into the grain until there is complete 

transformation of the grains to U4O9 [Einziger, 1992]. The grains remain in this phase for a 

temperature dependent duration until the fuel resumes oxidizing to the U3O8 state. The 

transformation to U3O8 occurs with ~33 percent lattice expansion that breaks the ceramic 

fragment structure into grain sized particles. At higher temperatures, the two transformations 

occur so rapidly that they are difficult to distinguish. The mechanism of oxidation in irradiated 

fuel appears to be different than in unirradiated fuel where U3O7 is formed and oxidation 

proceeds from the fragment surface and not down the grain boundaries. This mechanistic 

change occurs between ~10 and 30 Gwd/MTU. 

When the UO2 is in the form of a fuel rod, the expansion of the fuel, when it transforms to U3O8, 

induces a circumferential stress in the cladding. Due to the swelling of the fuel, the process is 

usually initially localized to the original cladding crack site. The cladding strains due to this 

stress range from 2-6 percent before the initial crack starts to propagate along the rod. The 

incubation time to initiate the propagation and the rate of propagation have an Arrhenius 

temperature dependence. Axial propagation, spiral propagation and a combination of the 

modes that result in splitting have been observed in PWR rods [Einziger, 1986]. 

8.7.2 Data Base 

The data base for oxidation was developed mostly in the 1980s in the US, Canada, England, 

and Germany. The data can usually appear in four forms: 1) O/M ratio (ratio of oxygen to metal 

content of the oxide) vs. time, 2) time to the UO2.4 plateau vs. time, 3) cladding splitting 

incubation vs. time, and 4) cladding splitting rate vs. time. Some later work was done by the 

Japanese on the effects of oxygen depletion [Nakamura, 1995], and most recently work is ongoing by the French primarily on MOX fuel. Much of the work was done on unirradiated fuel. All 
 8-42 

the work on cladding splitting was done in the early 1980s by the US [Einziger, 1984, 1986; 

Johnson, 1984] and Canadians [Novak, 1984; Boase, 1977] and is limited. Recently DOE 

[Bechtel, 2005] has issued an analysis of the oxidation issue in relationship to handling of 

potentially breached fuel in their proposed handling facility at the repository. This analysis 

depends on variables such as the gap between the fuel and the cladding, and burnup in a 

manner that is currently under technical review. In total, this research has shown that there are 

a number of variables that can affect the rates at which the fuel oxidizes and the cladding splits: 

burnup, moisture content of the air, cladding material, and type of initial defect. 

The DOE developed a model for fuel oxidation and cladding splitting [Bechtel, 2005] for use 

during long durations at the Yucca Mountain facility that tries to account for the fuel-to-cladding 

gap and burnup of the fuel. The gap is the as-measured cold gap and does not account for the 

closing of the gap due to differential thermal expansion of the cladding and fuel material, which 

could be calculated. There are inadequate data to verify correctness of the DOE model. Plots 

in the Einziger document [Einziger, 1986] present actual data and comparisons with the data 

taken by other researchers at 30 GWd/MTU. The gap closure is implicitly accounted for in the 

measurements of splitting. However, no burnup effects can be inferred from this data. 

No oxidation or cladding splitting studies have been conducted on fuel with burnup greater than 

45 GWd/MTU. Data between 30 and 45 GWd/MTU, shows a decrease in the oxidation rate due 

to the presence of certain actinides and fission products that are burned into the fuel. There is 

no reason that this should not continue at higher burnups, but the strength of the effect may 

change with burnup. Higher burnup fuel (>55 GWd/MTU) forms an external rim on the pellets 

that consists of very fine grains (1 micron). As indicated earlier, the oxidation process is a grain 

boundary effect. The fuel pellet must be divided into two regions for the purpose of oxidation 

analysis; the center of the pellet where the grains have grown slightly, and the rim. While the 

rate of the oxidation may decrease with burnup, the total amount of fuel that is oxidized may 

increase due to a much greater intergranular surface area in the rim region. The DOE model 

[Bechtel, 2005] uses a linear decrease in oxidation with burnup but this has, as yet, not been 

substantiated. A burnup effect is supported by Hanson’s analysis [Hanson, 1998] of Einziger 

and Cook’s data from the NRC whole-rod tests in which defect propagation was observed to 

occur earlier at the defects at the lower end of the rod where the burnup was lower. 

Studies using a low partial pressure of water vapor in air have not shown any dependence of 

the oxidation rate on the moisture content of the air [Ferry, 2005]. On the other hand, there are 

some studies that have shown a large increase in the oxidation rate when the moisture content 

is above 50 percent of the dew point. Oxidation in a 100 percent steam atmosphere is a 

different process. There are also studies that indicate that the oxidation rate will decrease if the 

oxygen content in the atmosphere drops into the range of a few torr or less [Nakamura, 1995]. 

It does not appear that there is an effect of oxygen content at higher oxygen levels but the data 

is sparse. 

Oxidation studies on fuel, with few exceptions, have been conducted on LWR fuel [Einziger, 

1986; Johnson, 1984]. However, the UO2 matrix is essentially the same in both PWR and BWR 

fuel. At the higher burnups, oxidation behavior may vary slightly as the actinide and fission 

product burn-in varies. The effect of the process on the splitting of the cladding may vary 

considerably due to the difference in gap size between the cladding types, and the thicker 

cladding in BWR rods. 

The limited cladding splitting studies have been conducted on Zircaloy clad PWR [Einziger, 

1984, 1986; Johnson, 1984] and CANDU fuel. Defects were put in the fuel either by an SCC 
 8-43 

(stress corrosion cracking) process producing small sharp holes more typical of those found in 

reactor initiated SCC and by drilling that produced a larger duller hole. Most of the defects used 

in the studies were of the latter type. No measurements were made in cladding above 

30 GWd/MTU. Very few data points were measured to determine the splitting rate; therefore, 

the time to start splitting has to be determined by interpolation. As a result, there is large 

uncertainty in both measurements. No measurements have been made on other alloy types 

(e.g., M5 and Zirlo) or at higher burnups where the cladding may be more brittle. Fuel oxidation 

would introduce uncertainties for fuel performance and fuel retrievability. 

8.8 Supplemental Information for Background justification for Cladding Temperature 

Considerations for the Storage of Spent Fuel (MEDIUM Priority) 

8.8.1 Basis for Guidance 

Creep is the dominant mechanism for cladding deformation under normal conditions of storage. 

The relatively high temperatures, differential pressures, and corresponding hoop stress on the 

cladding will result in permanent creep deformation of the cladding over time. Several 

laboratory programs have demonstrated that spent fuel has significant creep capacity even after 

15 years of dry cask storage. Einziger, et al., [2003] reported that irradiated Surry-2 PWR fuel 

rods (35.7 GWd/MTU) that were stored for 15 years at an initial temperature of 350°C (with 

temperatures reaching as high as 415°C for up to 72 hours) experienced thermal creep, which 

was estimated to be less than 0.1 percent. Post-storage creep tests were conducted to assess 

the residual creep capacity of the Surry-2 fuel rods. One-rod segment experienced a creep 

strain of 0.92 percent without rupture at 380°C and 220 MPa in 1820 hours (75.8 days). A 

different rod segment was tested at 400°C and 190 MPa for 1873 hours (78 days) followed by 

693 hours (28.9 days) at 400°C and 250 MPa, and experienced a creep strain of more than 

5 percent without failure [Tsai, 2002]. Profilometry measurements on that fuel rod indicated that 

the creep deformation was uniform around the circumference of the cladding with no signs of 

localized bulging, which can be a precursor for rupture. A report of the literature [Beyer, 2001] 

also indicates that some spent fuel cladding can accommodate creep strains of 2.87.5 percent 

at temperatures between 390 and 420°C and hoop stresses between 225 and 390 MPa. Other

significant contributions to the understanding of the effects of creep on spent fuel cladding can 

be found in several references [Einziger, et al., 1982; Rashid, et al., 2000; Hendricks, 2001; 

Rashid and Dunham, 2001; Machiels, 2002]. In general, these data and analyses support the 

conclusions that (1) deformation caused by creep will proceed slowly over time and will 

decrease the rod pressure, (2) the decreasing cladding temperature also decreases the hoop 

stress, and this too will slow the creep rate so that during later stages of dry storage, further 

creep deformation will become exceedingly small, and (3) in the unlikely event that a breach of 

the cladding due to creep occurs, it is believed that this will not result in gross rupture. 

Based on these conclusions, the staff has reasonable assurance that creep under normal 

conditions of storage will not cause gross rupture of the cladding and that the geometric 

configuration of the spent fuel will be preserved provided that the maximum cladding 

temperature does not exceed 400°C (752°F). As discussed below, this temperature will also 

limit the amount of radially oriented hydrides that may form under normal conditions of storage. 

The effects of normal conditions of storage (i.e., the decaying temperature and hoop stress on 

the cladding with time) can affect the metallurgical condition of spent fuel cladding containing 

significant amounts of hydrogen (e.g., spent fuel with high burnup levels). As the burnup level 

of the fuel increases beyond 45 GWd/MTU during reactor operation, the thickness of the oxide 
 8-44 

layer on the cladding increases. With increasing oxidation during reactor operation, the 

cladding absorbs more hydrogen. As discussed in Garde, et al., [1996], Chung and Kassner 

[1997], and Newman [1986], high burnup fuels tend to have relatively higher concentrations of 

hydrogen in the cladding. The hydrogen is present in the cladding predominantly as zirconium 

hydride precipitates, or particles. After the fuel is removed from the reactor, the zirconium 

hydrides are generally elongated and oriented circumferentially and are predominantly present 

in the outer rim of the cladding. At elevated temperatures, a percentage of the zirconium 

hydrides will dissolve, and under decreasing temperatures, zirconium hydrides will precipitate, 

or re-form. 

The materials phenomenon of hydride reorientation in zirconium-based alloys usually involves 

the dissolution of hydrides and the formation of zirconium-hydrides oriented perpendicular to the 

hoop stress (also referred to as radially oriented or radial hydrides) [Chung, 2000]. This occurs 

under sufficiently high hoop stresses along with the decrease in solubility of hydrogen that 

accompanies decreasing temperatures. The extent of the formation of radially oriented hydrides 

is a function of many parameters including the solubility of hydrogen in irradiated cladding 

material, cladding temperature, hoop stress, cooling rate, hydrogen concentration, thermal 

cycling, and materials characteristics. Among these parameters, the formation of radial 

hydrides is highly dependent on the hoop stress in the cladding. Data obtained from irradiated 

cladding [Einziger and Kohli, 1984; Cappelaere, et al., 2001; and, Goll, et al., 2001] indicate that 

stresses greater than 120 MPa (17.4 ksi) are required to initiate the formation of radial hydrides. 

Other data obtained from unirradiated zirconium-based cladding materials [Kese, 1998] indicate 

that radial hydrides can form at stresses as low as 90 MPa. Therefore, until the effects of 

reorientation are better understood, the hoop stress on the cladding should be controlled to 

preclude the formation of radially oriented hydrides. 

In general, a temperature limit of 400°C that is specified for normal conditions of storage and for 

short-term fuel loading and Part 72 storage operations (which includes drying, backfilling with 

inert gas, and transfer of the cask to the storage pad) will limit cladding hoop stresses and limit 

the amount of soluble hydrogen available to form radial hydrides. The use of a 400°C

temperature limit for normal conditions of storage and for short-term fuel loading and storage 

operations will simplify the calculations in SARs while assuring that hydride reorientation will be 

minimized. 

For low burnup fuel, a higher temperature limit may be used for short-term fuel loading and 

storage operations only, as long as the applicant can demonstrate that the best estimate 

cladding hoop stresses are equal to or less than 90 MPa for the temperature limit that is 

justified. For example, if the calculated best estimate hoop stress is equal to 90 MPa at 540°C, 

then 540°C is the maximum allowable temperature for loading operations. In this example, 

570°C is not the maximum allowable temperature limit. If the applicant can show that the best 

estimate hoop stress is less than or equal to 90 MPa at 570°C, then 570°C is the maximum 

allowable temperature. For some fuel types, short-term fuel loading and storage operation 

temperature limits as high as 570°C (1058°F) should be justified by the applicant. The materials 

reviewer should coordinate with the thermal reviewer to assure that either of the following 

criteria are used: (1) for low and high burnup fuel, the maximum calculated temperatures for 

normal conditions of storage and for fuel loading operations do not exceed 400°C, or (2) for low 

burnup fuel, a higher temperature limit may be used for loading and transfer operations, if the 

best estimate cladding hoop stress is less than 90 MPa for the temperature specified by the 

applicant. If the applicants use the latter approach, the materials reviewer should verify that the 

cladding hoop stresses are less than 90 MPa for each fuel assembly type (e.g., 14x14, 17x17, 

9x9, etc.) proposed for storage. Since the hoop stress is dependent on the rod internal 
 8-45 

pressure, cladding geometry, and the temperature of the gases inside the rod, the materials 

reviewer should coordinate with the thermal reviewer to verify that the applicant has calculated 

the best estimate hoop stress corresponding to the rod internal pressure of the highest burnup 

fuel assemblies of the specific type of assembly. It should be noted that during normal 

conditions of storage there will be a range of cladding temperatures that are less than the 

maximum allowable cladding temperature of 400°C, and this leads to a range of the internal rod 

pressures and the cladding hoop stresses, in any one storage cask. In general, the maximum 

allowable temperature will be 400°C or the maximum allowable temperature specified and 

supported (as discussed above) by the applicant. The maximum allowable temperature should 

be based upon the peak rod temperature, not the average rod temperature. By employing the 

peak rod temperature, only a small fraction of the rods will experience the temperature and 

stress conditions that could lead to the formation of radial hydrides during normal conditions of 

storage. 

It also has been observed and reported that thermal cycling (repeated heatup/cooldown cycles) 

can enhance the amount of hydrogen that eventually re-precipitates in the form of radial 

hydrides [Kammenzind, et al., 2000]. The extent of the formation of radial hydrides is 

dependent on many factors including the maximum temperature, change in temperature, 

number of thermal cycles, applied stress, hydrogen concentration, and solubility of hydrogen in 

the material. Kammenzind, et al., [2000] indicates that the formation of radial hydrides in spent 

fuel cladding can be minimized by restricting the change in cladding temperatures to less than 

65°C and minimizing the number of cycles to less than 10. The 65°C temperature limit is based 

upon the temperature drop required to obtain the degree of supersaturation required for the 

precipitation of hydrides in a short thermal cycle. 

For short-term accidents and short-term off-normal conditions that lead to an increase in 

temperature of the cladding, the dominant cladding failure mechanism is expected to be creep 

(stress rupture) of the cladding. To limit the amount of spent fuel that could be released from 

the cladding under off-normal conditions or accidents, the materials reviewer should coordinate 

with the thermal reviewer to verify that the maximum calculated cladding temperatures are 

maintained below 570°C (1058°F). The basis for using 570°C is established by the creep tests 

conducted on irradiated Zircaloy-4 rods [Einziger, et al., 1982]. The results from these 

experiments indicated that no cladding ruptures were observed for test times of 30 and 73 days. 

8.8.2 Review Guidance 

Prior to this guidance the short-term cladding temperature limit applicable to fuel loading 

operations was 570°C. All storage casks were certified using this limit. The current guidance to 

maintain cladding temperatures less than 400°C during fuel loading operations put into question 

whether the licensees who use certified storage casks (certified for fuel having average 

assembly burnups less than 45 GWd/MTU) would have to change their loading procedures and 

Technical Specifications to comply with this new temperature limit. Based on staff's evaluation, 

it is expected that fuel assemblies with burnups less than 45 GWd/MTU are not likely to have a 

significant amount of hydride reorientation due to limited hydride content. Further, most of the 

low burnup fuel has hoop stresses below 90 MPa. Even if hydride reorientation occurred during 

storage, the network of reoriented hydrides is not expected to be extensive enough in low 

burnup fuel to cause fuel rod failures. 

Given the conservatism used in calculating peak clad temperatures for low burnup fuel, the staff 

has reasonable assurance that storage cask systems which use the 570°C temperature limit for 

low burnup fuel loading operations will continue to perform as expected when the casks were 
 8-46 

originally certified. Therefore, there is no need to require the licensees of storage-only or dualpurpose cask systems to repackage spent fuel that was loaded using the 570°C temperature 

limit. Nevertheless, the 400°C limit is intended, with exceptions as stated above, to be generally 

applicable to all future loadings. Therefore, licensees are not required to modify their Technical 

Specifications or fuel loading procedures (i.e., vacuum drying) to meet the new 400°C limit for 

loading low burnup fuel into storage casks previously certified with the 570°C limit. Note that for 

future amendments to certified designs, the applicants may be required to comply with the 

400°C temperature limit as discussed above. 

8.9 Supplemental Information for the Design and Testing of Lid Welds on Austenitic 

Stainless Steel Canisters as Confinement Boundary for Spent Fuel Storage 

(MEDIUM Priority) 

8.9.1 Basis for the Review 

10 CFR 72.236(e) states: “The spent fuel storage cask [note: also called “canister”] must be 

designed to provide redundant sealing of confinement systems.” For a bolted lid canister 

design, the staff has accepted a dual seal arrangement as meeting the intent of this regulation. 

For a welded canister design, the staff has accepted closure designs employing redundant lids 

or covers, each with independent field welds. Thus, for either closure type, bolted or welded, a 

potential leak path must breach two independent seals or welds, sequentially, before the 

confinement system would be compromised. 

The construction codes specify the types of non-destructive examinations (NDE) required for 

the confinement boundary during canister fabrication and loading operations. In addition to the 

code required NDE, a helium leakage test of the confinement boundary is considered necessary 

to satisfy regulatory requirements. Whereas bolted lid canister designs incorporate a helium 

monitoring system during storage, the welded closure designs must rely on weld integrity to 

assure continued confinement effectiveness. Consequently, at least one of the redundant 

welded closures must be helium leak tested per the method of ANSI N 14.5, with one exception 

permitted. 

 

When the large, multi-pass weld joining the canister shell to the structural lid of an austenitic 

stainless steel spent fuel canister is executed and examined consistent with the guidance 

provided herein, the staff has reasonable assurance that no flaws of significant size will exist 

such that they could impair the structural strength or confinement capability of this weld. For a 

spent nuclear fuel canister, such a flaw would be the result of improper fabrication or welding 

technique, as service-induced flaws under normal and off-normal conditions of storage are not 

credible. Any such fabrication flaw would be reasonably detectable during the in-process and 

post-weld examination techniques described herein. 

Based on evaluation, these described techniques should detect any such flaw which could lead 

to a failure or credible leakage of radioactive material. Therefore, the staff believes that there is 

reasonable assurance that no credible leakage of radioactive material would occur through the 

structural lid to canister shell weld of an austenitic stainless steel canister, and that helium 

leakage testing of this specific weld is unnecessary provided the weld is executed and 

examined in accordance with the methods described herein. 

Conversely, it is the staff position that other welds associated with the lid assemblies of spent 

fuel canisters must be subject to the helium leak test of ANSI N 14.5, in addition to the ASME 
 8-47 

Code required pressure test and surface NDE in order to demonstrate compliance with 

10 CFR 72.236. 

Note the criteria outlined above do not supersede the flaw acceptance criteria of any 

construction code. Instead, this criteria is used to establish the maximum allowable weld 

deposit depth before an in-process penetrant test (PT) examination is required. 

8.9.2 Helium Leak Test 

The helium leak test was established to provide assurance that: 

 • No leakage occurred after the closure welds of the cask system were executed. 

This was viewed as necessary since no active or passive methods are employed 

to confirm or monitor the presence of helium within an all-welded spent fuel 

canister over its licensed lifetime. “No leakage” in this case means measured 

leak rate performed per ANSI N14.5, at a predetermined sensitivity that shows 

hypothetical doses would not exceed 10 CFR Part 72 limits. 

 • If the weld(s) meets the criteria of ANSI N14.5, the staff has assurance that radio 

nuclide leakage would not exceed the regulatory dose limits in 10 CFR Parts 

72.104 and 72.106. 

 • No oxygen in-leakage could occur, thereby assuring the presence of the inert 

helium atmosphere which prevents oxidation and corrosion induced degradation 

of the spent fuel assemblies and enhances cooling of the spent fuel. 

Helium Leak-Testing of the Confinement Boundary 

The redundant weld requirement for the confinement system closure creates two closure 

boundaries. The staff should verify that at least one of the redundant boundaries is helium leak 

tested, or, some closure welds leak tested and the remaining closure welds of the same 

boundary designed so that the “large weld” exemption criteria of this guidance are met. Only a 

boundary which is testable or excluded from testing, per this guidance, should be considered 

the confinement boundary of the redundant closures. Refer to Figures 8-3 and 8-9 and the 

following narrative for application of this criteria to two currently approved designs: 

Leak Testing a Single Lid With Cover Plate Design – Figure 8-3 

In Figure 8-3, the dotted line marked (1) defines one closure boundary. Starting on the left side 

of the sketch, the closure boundary can be traced from the canister wall, up through the large, 

multi-pass weld joining the canister wall to the heavy section, combined shield and structural lid. 

The boundary continues through the lid to the small weld joining the heavy lid to the vent-anddrain port closure plate, and back to the heavy lid again. The remainder of the boundary (and 

sketch) is assumed to be symmetrical with or similar to the half-sketch portion that is shown, for 

all cases. 

This boundary demonstrates confinement integrity by means of the large weld exemption 

criteria for one weld and by helium leak testing the small cover plate weld. 

The large, canister-shell-to-lid weld is exempted from the helium leak test. This is because the 

canister shell to lid weld is a large, multi-pass weld meeting the flaw tolerance and other 
 8-48 

appropriate portions of this guidance. Note that this weld is executed prior to filling the canister 

with helium (excluding purging/welding gas). 

Before the remaining welds of this first closure boundary are executed, the canister is drained, 

dried, purged, and filled with helium to the design operating pressure. The helium line

connection is closed off and the cover plate fitted and welded into place. Since the cover plate 

weld may have potentially been pressurized from underneath due to assumed leakage from the 

closure valve, it must be helium leak tested in accordance with the methods described in ANSI 

N14.5-1997. If there are other cover plates and welds, they would also be helium leak tested. 

This completes the first closure boundary. Note again that one weld was exempted from the 

helium leak test by the design criteria. The other weld was leak tested. Thus, this closure 

boundary demonstrates compliance with regulatory requirements and is consistent with the staff 

guidance by ensuring at least one of the two redundant closure boundaries is leak tested or 

exempted from leak testing by conformance with the large-weld exemption guidance. This 

boundary thus also qualifies as the confinement boundary. 

The second boundary, delineated by line 2, can be traced from the canister wall on the left side 

of the sketch up through the cover plate fillet weld joining the canister wall to the structural lid 

cover plate. The boundary continues through the cover plate to the fillet weld joining the cover 

plate to the canister lid. The weld joining the cover plate to the canister wall and lid cannot be 

helium leak tested since there is no feasible means to do so. However, since the first closure 

boundary, delineated by line 1, was tested (or exempted thru design), the need to helium leak 

test at least one of the closure boundaries has been satisfied. Since this second boundary does 

not meet all the criteria for a confinement boundary, it may not be designated as the 

confinement boundary. The first closure is thereby the confinement boundary in this design, as 

it meets all the applicable criteria for a confinement boundary. 

Leak Testing a Dual Lid Design – Figure 8-4 

In Figure 8-4 of this SRP, the dotted line marked (1) defines one of the redundant closure 

boundaries. It may be traced from the canister wall on the left side of the sketch. The boundary 

proceeds through the partial penetration weld joining the canister wall to the shield lid and into 

the shield lid. It continues through the small fillet weld joining the vent/drain port cover plate, the 

cover plate, and back through the same fillet weld to the shield lid. 

This closure boundary may satisfy the leak test guidance by several methods, depending on 

details of the weld design. The canister shell to shield lid weld may be designed several ways. 

The weld may be a small seal weld which would necessitate subsequent helium leak testing. 

Conversely, it could be a large, multi-pass weld consistent with the guidance described herein. 

In that case, the weld would qualify for the leak test exemption. Either way, note that this weld 

(canister to shield lid weld) is executed prior to filling and pressurizing the canister with helium 

(use of purge or backing gas for welding operations is not considered filling or pressurizing). 

Next, the canister is drained, dried, purged, and filled with helium to the design operating 

pressure. The helium line connection is closed off. The cover plate is fitted and welded into 

place. Since this weld may potentially be pressurized from underneath due to assumed leakage 

through the closure valve, it must be helium leak tested regardless of weld size (thickness). 

This completes the first closure boundary. Note that one weld was either tested, or, exempted 

from the helium leak test by the design criteria. The other weld was leak tested. Thus, this 

closure boundary demonstrates compliance with regulatory requirements and is consistent with 
 8-49 

Canister 

Figure 8-3 Single Lid with Cover Plate Design 

1

2

Cover Plate

Vent or Drain Port Cover Plate

2 1

Combined Shield and Structural Lid

 

staff guidance by ensuring at least one of the two redundant closures is leak tested or exempted

by conformance to this guidance. This closure may therefore be designated as the confinement 

boundary. 

The secondary boundary, delineated by line 2 in sketch B, can be traced from the canister wall 

on the left side of the sketch up through the canister wall-to-structural lid weld and into the 

structural lid. 

The weld joining the canister wall and structural lid cannot be helium leak tested because 

helium is not present. Note, however, that this weld complies by design with the criteria 

described herein due to its size, structural requirements and weld examination requirements of 

the governing construction code. 

In this case, the second closure also qualifies for designation as the confinement boundary 

because the single large weld involved may be exempted from the helium leak test. In this 

design, the designer therefore has the freedom to designate either of the redundant closures as 

the confinement boundary. Only one of the two closures is designated as the confinement 

boundary. 
 8-50 

Figure 8-4 Dual Lid Design 

2

1 

2 1

Canister Structural Lid

Vent or Drain Port Cover Plate

Shield 

Lid 

Shield

Lid 
 9-1 

 9 OPERATING PROCEDURES EVALUATION 

9.1 Review Objective 

The operating procedures review ensures that the applicant's safety analysis report (SAR) 

presents acceptable operating sequences, guidance, and generic procedures for the key 

operations shown in Section 9.2, “Areas of Review.” The review also ensures that the SAR 

incorporates and is compatible with the applicable operating control limits in the technical 

specifications. 

The operating sequences described in the SAR should provide an effective basis for the 

development of the more detailed operating and test procedures by the cask user when 

preparing and implementing detailed site-specific procedures. The NRC normally inspects 

selected site-specific procedures. Such procedures are important aspects of the site’s radiation 

protection program and allow the cask user to safely store spent nuclear fuel (SNF). 

This chapter applies to all discipline reviews. Figure 1-1 presents an overview of the evaluation 

process and can be used as a guide to assist in coordinating with other review disciplines. 

9.2 Areas of Review 

This chapter of the dry storage system (DSS) Standard Review Plan (SRP) provides guidance 

in evaluating the applicant’s general operating sequences and generic procedures related to 

cask operations. Within each area of cask operations, the NRC staff assesses the effectiveness 

of the applicant’s generic procedures on a technical and safety basis for the subsequent 

development of detailed operating procedures. As required by U.S. Code of Federal 

Regulations (CFR) Part 72, “Licensing Requirements for the Independent Storage of Spent 

Nuclear Fuel and High-Level Radioactive Waste,” Title 10, “Energy” (10 CFR Part 72) 72.234(f), 

these procedures are to be provided to each cask user for the subsequent preparation and 

implementation of detailed site-specific procedures by the cask system user acting under a 

general license. Areas of review addressed in this chapter include the following: 

Loading Operations 

 Fuel Specifications 

 Damaged Fuel 

 Subcriticality Features 

 ALARA 

 Offsite Release 

 Draining and Drying 

 Filling and Pressurization 

 Welding and Sealing 

 Administrative Programs 

Cask Handling and Storage Operations

Cask Unloading 

 Damaged Fuel 

 Cooling, Venting, and Reflooding 

 Fuel Crud 

 ALARA 

 Offsite Release 
 9-2 

9.3 Regulatory Requirements 

This section presents a summary matrix of the portions of 10 CFR Part 72 that are relevant to 

the review areas addressed by this chapter. The NRC staff reviewer should read the exact 

referenced regulatory language. Table 9-1 matches the relevant regulatory requirements 

associated with this chapter to the areas of review. 

Table 9-1 Relationship of Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Regulations

72.104 

(b), (c) 

72.122(f), 

(h)(1), (l) 

72.212 

(b) (9) 

72.234 

(f) 

72.236 

(c) 

72.236 

(h), (i) 

Cask Loading 

Operations ● ● ● ● ●

Cask Handling and 

Storage Operations ● ● ● ● ●

Cask Unloading ● ● ● ● ●

9.4 Acceptance Criteria 

Chapter 9, “Operating Procedures Evaluation,” of the SAR should identify and describe the 

sequence of significant operations and actions that are important to safety for cask loading, 

cask handling, storage operations, and cask unloading. A sufficient level of detail is needed in 

Chapter 9 of the SAR for the reviewer to conclude that operating procedures will adequately 

protect health and minimize danger to life or property, protect the fuel from significant damage 

or degradation, and provide for the safe performance of tasks and DSS operations. 

This portion of the DSS review seeks to ensure that the generic procedure descriptions and 

operational sequences described in the SAR include the following information: 

 • Major operating procedures should apply to the principal activities expected to 

occur during dry storage. The expected scope of activities for the SAR operating 

procedure descriptions is previously described in Section 9.2 as well as Chapter 

8 of Regulatory Guide (RG) 3.61, “Standard Format and Content for a Topical 

Safety Analysis Report for a Spent Fuel Dry Storage Cask.” Operating 

procedure descriptions should be submitted to address the cask design features 

and planned operations. 

 • Operating procedure descriptions should identify measures to control processes 

and mitigate potential hazards that may be present during planned normal 

operations. Section 9.5, “Review Procedures,” in this chapter discusses 

previously identified processes and potential hazards. 
 9-3 

 • Operating procedure descriptions should ensure conformance with the applicable 

operating controls and limits described in the cask system’s Technical 

Specifications provided in Chapter 13, “Technical Specifications and Operating 

Controls and Limits Evaluation,” of the SAR. 

 • Operating procedure descriptions should reflect planning to ensure that 

operations will fulfill the following acceptance criteria: 

- Occupational radiation exposures will remain as low as is reasonably 

achievable (ALARA). 

- Effective measures will be taken to preclude potential unplanned and 

uncontrolled releases of radioactive materials. 

- Offsite dose rates will be maintained within the limits of 10 CFR Part 20 

and 10 CFR 72.104 for normal operations, and 10 CFR 72.106 for 

accident-level conditions. 

In addition, the operating procedure descriptions should support and be 

consistent with the bases used to estimate radiation exposures and total doses 

as defined in Chapter 11, “Radiation Protection Evaluation,” of this SRP. 

 • Operating procedure descriptions should include provisions for the following 

activities: 

- Testing, surveillance, and monitoring of the stored material and casks 

during storage and loading and unloading operations. 

 - Contingency actions triggered by inspections, checks, observations, 

instrument readings, and so forth. Some of these may involve off-normal 

conditions addressed in Chapter 12, “Accident Analyses Evaluation,” of 

the SAR. 

9.4.1 Cask Loading 

In addition to the acceptance criteria above, additional acceptance criteria for cask loading are 

as follows: 

 • The operating procedure descriptions should facilitate reducing the amount of 

water vapor and oxidizing material within the confinement cask to an acceptable 

level to protect the SNF cladding against degradation that might otherwise lead 

to gross ruptures. 

 • Operating procedures should specify methods for placing damaged fuel in a 

damaged-fuel can prior to loading into a cask, if applicable. 

9.4.2 Cask Handling and Storage Operations 

In addition to the acceptance criteria stated above, operating procedure descriptions should 

include provisions for maintenance of casks and cask functions during storage. 
 9-4 

9.4.3 Cask Unloading 

In addition to the acceptance criteria stated above, operating procedures should facilitate ready 

retrieval of SNF stored in a storage cask. 

9.5 Review Procedures 

Introduction (MEDIUM Priority)

The interrelationship of the operating procedures evaluation with other disciplines is shown in 

Figure 9-1. 

The review procedures described in this section are presented in a format intended to facilitate 

an independent review. Even though several individuals may actually be tasked with preparing 

the chapter of the safety evaluation report (SER) related to operating procedures, all review 

team members should examine the operating procedure descriptions presented in the SAR. If 

the descriptions included in the SAR are not sufficiently detailed to allow a complete evaluation 

concerning fulfillment of the acceptance criteria, reviewers should request additional information 

from the applicant. 

The operating procedure sequences are described in Chapter 9 of the SAR, and the direct dose 

rate information in Chapter 6, “Shielding Evaluation,” of the SAR is used to assess compliance 

with radiation protection requirements in Chapter 11 of the SAR. The reviewer should verify that 

the evaluation of Chapter 9 of the SAR is coordinated with the shielding and radiation protection 

evaluations covered in Chapters 6, “Shielding Evaluation” and 11, “Radiation Protection 

Evaluation,” of this SRP. 

In addition, the following review procedures are based on the assumption that the ISFSI 

operations are at a reactor facility licensed under 10 CFR Part 50, “Domestic Licensing of 

Production and Utilization Facilities,” and that loading and unloading activities will be performed 

in the facility’s SNF pool. Review procedures for dry fuel transfers and/or ISFSI operations at 

sites away from a reactor will be developed at a later date, if necessary. 

Reviewers should be familiar with ANSI/ANS 57.9, “Design Criteria for an Independent Spent 

Fuel Storage Installation (Dry Type),” which applies to DSS operating procedures. Background 

information is available in NUREG/CR-4775, “Guide for Preparing Operating Procedures for 

Shipping Packages,” which provides guidance on preparing operating procedures for shipping 

packages. Although NUREG/CR-4775 specifically addresses 10 CFR Part 71, most of the 

guidance can be adapted for storage casks that are governed by 10 CFR Part 72. 

Consequently, reviewers should be familiar with this information before initiating the DSS 

operating procedures review. 

Since many of the detailed procedures may be developed by facilities licensed under 10 CFR 

Part 50 or 72, further background information on site-specific procedure requirements may be 

found in RG 1.33, “Quality Assurance Program Requirements (Operation),” and its associated 

standard ANSI/ANS 3.2. Reviewers of Chapter 9, “Operating procedures Evaluation” of the 

SAR should also be familiar with Chapter 11, “Conduct of Operations Evaluation,” of NUREG1567, “Standard Review Plan for Spent Fuel Dry Storage Facilities.” Specifically, Section 

11.4.3, “Normal Operations,” in NUREG-1567 provides NRC review acceptance criteria for 

facility-developed procedures. 
 9-5 

Figure 9-1 Overview of Operating Procedures Evaluation 

Chapter 2 – Principal 

Design Criteria Evaluation

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation

• Ready Retrieval of 

Contents

• Waste Minimization

• Decontamination

• Maintenance Procedures

Chapter 1 – General 

Information Evaluation

• General Restrictions

Cask Unloading

• Damaged Fuel

• Cooling, Venting, and 

Reflooding

• Fuel Crud

• ALARA

• Offsite Release

Loading Operations Cask Handling and Storage Operations

• Fuel Specifications

• Damaged Fuel

• ALARA

• Offsite Release

• Cask Draining and Drying

• Cask Filling and 

Preservation

• Cask Welding and Sealing

• Subcriticality Features

Chapter 9 – Operating Procedures Evaluation

Cask Unloading

• Damaged Fuel

• Cooling, Venting, and 

Reflooding

• Fuel Crud

• ALARA

• Offsite Release

Loading Operations Cask Handling and Storage Operations

• Fuel Specifications

• Damaged Fuel

• ALARA

• Offsite Release

• Cask Draining and Drying

• Cask Filling and 

Preservation

• Cask Welding and Sealing

• Subcriticality Features

Chapter 9 – Operating Procedures Evaluation

Chapter 3 – Structural 

Evaluation

• Closure Requirements

• Tie -down Requirements

• Handling Restrictions

Chapter 4 – Thermal 

Evaluation

• Subcooling Margin

• Cask Reflooding and 

Cooldown Quench 

Temperature and Flow 

Rates

Chapter 14 – Quality 

Assurance Evaluation

• Procedures Affecting 

Quality

• Controls on Special 

Processes

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation

• Operational Limits

• Visual and Nondestructive 

Examination Inspection

• Cask Identification

• Repair, Replacement, and 

Maintenance

Chapter 6 – Shielding 

Evaluation

• Radiation Limits

Chapter 7 – Criticality 

Evaluation

• Fuel Specifications

• Soluble Neutron Poison

• Fuel Configuration

• Subcriticality Features

Chapter 8 – Materials 

Evaluation

• Welded Lids

• Bolt Applications

Chapter 14 – Quality 

Assurance Evaluation

• Tests

• Corrective Actions

Chapter 11 – Radiation 

Protection Evaluation

• ALARA

• Radiation Levels

• Environmental Protection 

Methods

Chapter 12 – Accident 

Analyses Evaluation

• Corrective Actions

Chapter 13 – Tech. Specs. 

And Operating Controls 

Evaluation

• Surveillance 

Requirements

• Administrative Controls

• Monitoring Instruments

• Functional/Operating 

Limits

Chapter 5 – Confinement 

Evaluation

• Monitoring
 9-6 

In general, reviewers should perform the following steps in the process of evaluating all of the 

operating procedure descriptions and operational sequences provided in the SAR. 

• Verify that the proposed operating procedure descriptions incorporate and are 

compatible with the applicable operating limits and controls in Chapter 13, “Technical 

Specifications and Operational Controls and Limits Evaluation” of the SAR. 

Coordinate with the review of operating controls and limits, as described in Chapter 

13, “Technical Specifications and Operating Controls and Limits Evaluation,” of this 

SRP. 

• Ensure that the proposed operating procedure descriptions properly consider the 

prevention of hydrogen gas generation from any cause (including the reaction of zinc 

primer coating with acidic pool water, radiolysis, or other causes). Prevention of 

hydrogen generation or adequate purging of hydrogen is essential during loading 

and unloading operations that involve seal welding, seal cutting, grinding, or other 

forms of hot work. 

• Determine whether the descriptions include appropriate precautions to minimize 

occupational radiation exposures in accordance with ALARA principles and the limits 

given in 10 CFR Part 20, as mandated by 72.126(a)(5). Provisions may include use 

of remotely controlled equipment, monitoring, and use of portable shielding. 

• Verify that the operating procedure descriptions include a general listing of the major 

tools and equipment needed to support ISFSI loading, storage, and unloading 

operations (including those at the pool facility). The descriptions should also address 

installation, use, and removal of the cask and fuel, tools, and equipment. In addition, 

the descriptions should describe any specialized tools and equipment in sufficient 

detail to enable users to understand their function. Examples include lifting yokes, 

transporter equipment, welding and cutting equipment, and vacuum drying 

equipment. The use of any such equipment that is classified as being important to 

safety is subject to approval as part of the application review. Such equipment 

should be identified and described in detail, its performance characteristics should be 

defined, and the design should be evaluated. 

In addition to these generic review procedures, all disciplines should evaluate each of the 

specific areas of operating procedure review as described in the following subsections. 

9.5.1 Cask Loading (Priority - as indicated) 

(MEDIUM Priority) The operating procedure descriptions in the SAR should present the 

activities sequentially in the anticipated order of performance. The generic procedures in 

Chapter 9, “Operating Procedures Evaluation” of the SAR should be reviewed to ensure that 

they include appropriate key prerequisite, preparation, and receipt inspection activities to be 

accomplished before cask loading. The reviewer should verify that tests, inspections, 

verifications, and cleaning procedures required in preparation for cask loading are specified. In 

addition, where applicable, the reviewer should verify that the procedure descriptions include 

actions needed to ensure that any fluids such as shield water and primary coolants fill their 

respective cavities according to design specifications. 
 9-7 

Fuel Specifications (MEDIUM Priority) 

The reviewer should verify that the loading procedure description appropriately addresses the 

SNF specifications (e.g., burnup, cooling period, source terms, heat generation, cladding 

damage, associated non-fuel hardware, etc.) in Chapter 2, “Principal Design Criteria,” and 

Chapter 13, “Technical Specifications and Operation Controls and Limits Evaluation” of the 

SAR. For cask systems relying upon burnup credit, the loading procedure description should 

include verification that assemblies selected for loading meet the specifications for assembly 

operational history and the loading curve as well as include performance of measurements to 

confirm assembly burnup values. Depending on the types and specifications of fuel assemblies 

stored in the reactor SNF pool, detailed site-specific procedures may be necessary to ensure 

that all fuel loaded in the cask meets the fuel specifications for the cask design. These 

procedures can be evaluated only on a site-specific basis and will generally be evaluated 

through inspections rather than during the licensing review. The SAR should indicate, however, 

that such procedures may be necessary. 

Damaged Fuel (MEDIUM Priority) 

The reviewer should verify that the SAR includes appropriate measures for the loading of 

damaged fuel, if damaged fuel is included in the proposed cask contents. Chapter 2, “Principal 

Design Criteria Evaluation,” and Chapter 8, “Materials Evaluation,” of this SRP provide criteria 

for the storage of damaged fuel. Information in Section 8.6, “Supplemental Information for 

Methods for Classifying Fuel,” of this SRP should be used to identify the conditions that 

determine when SNF is to be classified as damaged fuel. Section 8.4.17.2 of this SRP should 

be reviewed to determine the classification, documentation, and special handling requirements 

for damaged fuel and determine if operating procedures address these requirements. 

Subcriticality Features (MEDIUM Priority) 

Where applicable, the reviewer should verify that the procedure descriptions include the use of 

features important to criticality safety that may require installation by the DSS user. Such items 

include fuel spacers and items (e.g., blocks) used to prevent loading of contents in selected 

basket locations. The procedure descriptions should include installation, or verification of the 

installation, of these items prior to cask loading for casks that rely upon these features in the 

criticality analysis. Additionally, the procedure descriptions should include verification, in 

accordance with Technical Specification requirements, of the minimum soluble boron level 

necessary for cask loading for casks requiring soluble boron to meet subcriticality. 

ALARA (LOW Priority) 

The reviewer should verify that the procedure descriptions incorporate ALARA principles and 

practices. These may include provisions to perform radiological surveys as well as exposure 

and contamination control measures, temporary shielding, and suggested caution statements 

related to actions that could change radiological conditions. In addition, the reviewer should 

verify that any recommended surveys incorporate the applicable operating controls and limits 

described in Chapter 13, “Technical Specifications and Operating Controls and Limits 

Evaluation” of the SAR. 

Offsite Release (LOW Priority) 
 9-8 

Where applicable, the reviewer should verify that the SAR describes methods to minimize offsite 

releases such as decontamination, filtered ventilation, temporary containments (tents), and so 

forth. The procedure descriptions should also provide for minimizing generation of radioactive 

waste. 

Draining and Drying (MEDIUM Priority) 

The reviewer should evaluate the descriptions related to methods for use in draining and drying 

the cask for ISFSI operations at a reactor facility or at sites away from a reactor with a transfer 

pool. In particular, the descriptions should clearly describe the procedures for removing water 

vapor and oxidizing material to an acceptable level, and the reviewer should assess whether 

those procedures are appropriate. 

The NRC staff has accepted vacuum drying methods comparable to those recommended in 

PNL-6365 (Knoll, 1987). This report evaluates the effects of oxidizing impurities on the dry 

storage of light-water reactor (LWR) fuel and recommends limiting the maximum quantity of 

oxidizing gasses (such as O2, CO2

4

, and CO) to a total of 1 gram-mole per cask. This 

corresponds to a concentration of 0.25 volume percent of the total gases for a 7.0m3

 (about 

247 ft3

) cask gas volume at a pressure of about 0.15 MPa (1.5 atm) at 300°K (80.3°F). This 

1 gram-mole limit reduces the amount of oxidants below levels where any cladding degradation 

is expected. Moisture removal is inherent in the vacuum drying process, and levels at or below 

those evaluated in PNL-6365 (about 0.43 gram-mole H2O) are expected if adequate vacuum 

drying is performed. 

If alternative methods other than vacuum drying are used (such as forced helium recirculation), 

the reviewer should ensure that additional analyses or tests are provided to sufficiently justify 

that cover gas moisture and impurity levels as specified in Chapter 9, “Operating Procedures 

Evaluation” of the SAR are met and will not result in unacceptable cladding degradation. 

The following examples illustrate the accepted methods for cask draining and drying in 

accordance with the recommendations of PNL-6365 (Knoll, 1987): 

 • The cask should be drained of as much water as practicable and evacuated to 

less than or equal to 4.0E-04 MPa (4 millibar, 3.0 mm Hg or Torr). After 

evacuation, adequate moisture removal should be verified by maintaining a 

constant pressure over a period of about 30 minutes without vacuum pump 

operation (or the vacuum pump is running but it is isolated from the cask with its 

suction vented to atmosphere). The cask is then backfilled with an inert gas 

(e.g., helium) for applicable pressure and leak testing. Care should be taken to 

preserve the purity of the cover gas and, after backfilling, cover gas purity should 

be verified by sampling. 

 • The procedures should reflect the potential for blockage of the evacuation 

system or masking of defects in the cladding of non-intact rods, as a result of 

icing during evacuation. Icing can occur from the cooling effects of water 

vaporization and system depressurization during evacuation. Icing is more likely 

to occur in the evacuation system lines than in the cask because of decay heat 

from the fuel. A staged draw down or other means of preventing ice blockage of 

 

4

 Can be broken down by radiolysis. 
 9-9 

the cask evacuation path may be used (e.g., measurement of cask pressure not 

involving the line through which the cask is evacuated). 

 • The procedures should specify a suitable inert cover gas (such as helium) with a 

quality specification that ensures a known maximum percentage of impurities to 

minimize the source of potentially oxidizing impurity gases and vapors and 

adequately remove contaminants from the cask. 

 • The process should provide for repetition of the evacuation and repressurization 

cycles if the cask interior is opened to an oxidizing atmosphere following the 

evacuation and repressurization cycles (as may occur in conjunction with 

remedial welding, seal repairs, etc.). 

Reviewers should ensure that the drying specifications are consistent with the proposed 

operating controls and limits described in the technical specifications provided in Chapter 13 of 

the SAR. In addition, reviewers should assess the need for any additional technical 

specifications. 

Welding and Sealing (HIGH Priority) 

Structural and materials disciplines should coordinate their review of welded lids as described in 

Section 8.4.7, “Weld Design/Inspection,” of this SRP for application of the proper weld joint, 

welding procedures, and non-destructive examination methods (NDE) to ensure the appropriate 

operating procedures are in place and acceptable. Reviewers should verify that procedures are 

acceptable for NDE and welding of the closure welds. While the NRC accepts progressive 

surface examinations utilizing dye penetrant testing (PT) or magnetic particle (MT) examination, 

it is only permitted if unusual design or loading conditions exist. In addition, if a PT or MT 

examination is used, a stress-reduction-factor of 0.8 is imposed on the weld strength for the 

reasons presented in Section 8.4.7.3. The SAR should also ensure ALARA principles are 

followed and include acceptable provisions for correcting weld defects and any additional drying 

and purging that may be necessary. 

The reviewer should verify that provisions for placing and tightening any closure bolts, such as 

those associated with concrete casks, are consistent with information presented in Chapters 2, 

3, and 10 of the SAR that address applicable design criteria, structural evaluation, and the 

acceptance tests and maintenance program, respectively. The materials discipline should 

ensure that the closure bolts satisfy the conditions given in Section 8.4.10, “Bolt Applications,” of 

this SRP. The SAR should specify the torque required to properly seal the closure lid. The 

inner seal should be tested using a helium leak test with the interior of the cask pressurized as 

previously described. The outer seal should also be tested using a helium leak test with the 

between-seal volume pressurized as required by the respective subsection of the ASME B&PV 

Code, Section III. 

Filling and Pressurization (LOW Priority) 

The reviewer should verify that the procedure recommendations address steps to fill and 

pressurize the cask with inert gas such as helium with a known maximum percentage of 

impurities. The operating procedures should state that the filling and pressurization (or 

evacuation and backfill) process be repeated if the cask cavity is exposed to the atmosphere. 

Also, the reviewer should ensure that the procedure recommendations include the requirements 

in Chapter 13, “Technical Specifications and Operation Controls and Limits Evaluation” of the 

SAR. 
 9-10 

The SAR should specify the leak rate criteria (e.g., total leakage, leakage per closure, 

sensitivities of tests, etc.), and the reviewer should verify that these criteria are consistent with 

those presented in Chapters 2, 9, and 13 of the SAR. In addition, the reviewer should assess 

the general methods of leak testing (e.g., pressure rise, mass spectrometry) as they apply to the 

leak rate being tested. Particular attention should be paid to the possible use of quickdisconnect fittings for draining and filling operations. Although no credit is usually taken for 

these devices as part of the confinement boundary, their presence can negate the results of the 

leak test, and the SAR should provide guidance regarding their use. In addition, the guidelines 

presented in the SAR should note that leak testing is in accordance with ANSI N14.5, 

“Radioactive Materials – Leakage Tests on Packages for Shipment.” 

The reviewer should ensure that the SAR presents applicable pressure testing criteria (e.g., test 

pressure, hold periods, inspections) and that these criteria are consistent with those presented 

in Chapter 9 of the SAR. 

Administrative Programs (HIGH Priority) 

The applicant may request that one or more administrative programs be approved by the NRC 

in lieu of the requirements set forth in Section 9.5.1 above for offsite releases, draining and 

drying, filling and pressurization, and welding and sealing. Requirements for such 

administrative programs are provided in NUREG-1745, “Standard Format and Content for 

Technical Specifications for 10 CFR Part 72 Cask Certificates of Compliance,” and are 

summarized in this section. 

The applicant may request the NRC approve an administrative program for offsite releases. In 

this case, the reviewer should verify that the SAR describes a Radioactive Effluent Control 

Program and related operating procedures that shall be established, implemented, and 

maintained to: 

 • Implement the requirements of 10 CFR 72.126. 

 • Limit the surface contamination and verification of meeting those limits prior to 

removal of the cask from the Part 50 structure. 

 • Limit the leakage rate and verification of meeting those limits prior to removal of 

the cask from the Part 50 structure. 

 • Show compliance with the requirements of 10 CFR 72.104 and 72.106. 

In addition, the applicant may request the NRC approve an administrative program for cask 

loading. In this case, the reviewer should verify that the SAR requirements are implemented for 

loading fuel and components into the cask and preparing the cask for storage. The 

requirements of the program for loading and preparing the cask should be completed prior to 

removing the cask from the 10 CFR Part 50 structure. (Items 1, 5, and 6 below are associated 

with requirements that will remain in the technical specifications; however, the process for 

establishing the specified action limit may be moved to this administrative program if a method 

of evaluation acceptable to the NRC is presented in the SAR. Items 2, 3, and 4 have been 

relocated from the Limiting Conditions of Operations [LCO] section to this administrative 

program because it is felt that NRC-approved methods of evaluation will be relatively easy to 

develop. If appropriate methods are not presented in the SAR, these items will retain LCOs.) 
 9-11 

At a minimum, the cask-loading program shall establish criteria that need to be verified to 

address SAR commitments and regulatory requirements for: 

 1. Vacuum drying times and pressures, or forced helium drying criteria,, to assure 

that the short-term fuel temperature limits are not violated and the cask is 

adequately dry. 

 2. Inerting pressure and purity to assure adequate heat transfer and corrosion 

control. 

 3. Leak testing to assure adequate cask integrity and consistency with the offsite 

dose analysis. 

 4. Surface dose rates to identify significant problems with shielding fabrication, 

gross misloads, and verify consistency with the offsite dose analysis. 

 5. Ambient and pool water temperature to assure adequate subcriticality and 

material ductility. 

 6. SNF pool boron concentration to verify the acceptable subcriticality margin. 

 7. Clad oxidation thickness for high-burnup fuel in accordance with SRP Chapter 8, 

:Materials Evaluation” or other NRC-approved methodology if high-burnup fuel is 

included in the contents. 

The program shall include compensatory measures and appropriate completion times if the 

program requirements are not met. 

9.5.2 Cask Handling and Storage Operations (LOW Priority) 

The reviewer should examine the recommendations associated with procedures necessary to 

transfer the cask to the storage location. The reviewer should pay particular attention to 

ensuring that all accident events applicable to such transfer are bounded by the design events 

analyzed in Chapters 2, “Principal design Criteria”, 3, “Structural Evaluation” and 12, “Accident 

Analyses Evaluation” of the SAR. This includes procedures to be specified in the SAR for use 

after a design-basis accident for testing the effectiveness of the shielding. The structural and 

thermal disciplines should coordinate their review to ensure that all conditions for lifting and 

handling methods are bounded by the evaluations in their respective Chapters 3 and 4 of the 

SAR. There may be technical specifications associated with cask transfer operations such as 

restricting lift heights and environmental conditions (e.g., high/low temperatures, etc.) requiring 

coordination with the review in Chapter 13, “Technical Specifications and Operating Controls 

and Limits Evaluation,” of this SRP. 

The reviewer should verify that the procedure recommendations discuss the inspection, 

surveillance, and maintenance requirements that are applicable during ISFSI storage. 

Surveillance and monitoring requirements should also be included in Chapter 13 of the SAR, 

and maintenance should be included in Chapter 10 of the SAR. Reviewers should note that if 

the confinement vessel closure is bolted, the NRC staff generally requires that the successful 

operation of the seals be demonstrated with an initial leak test and a monitoring system and/or a 
 9-12 

surveillance program as discussed in Chapter 10, “Acceptance Tests and Maintenance Program 

Evaluation,” of this SRP. 

The shielding and radiation protection reviewers should verify that proposed procedures give 

due consideration to maintaining doses ALARA during cask handling and storage operations. 

The applicant may request that an ISFSI Operations Program be approved by the NRC. 

Requirements for such an administrative program are provided in NUREG-1745. The reviewer 

should verify that such a program establishes criteria for: 

 • Minimum cask center-to-center spacing. 

 • Pad parameters (i.e., pad thickness, concrete strength, soil modulus, 

reinforcement, etc.) that are consistent with the SAR analysis. 

 • Maximum lifting heights for the cask system to ensure that the gravity load limits 

are met for the design-basis events. 

9.5.3 Cask Unloading (Priority – as indicated) 

(LOW Priority) The reviewer should verify that the SAR adequately describes the necessary 

unloading procedure recommendations. The unloading procedure descriptions should present 

the activities sequentially in the anticipated order of performance, including those key 

prerequisite and preparation tasks that must be accomplished before cask unloading. Where 

applicable, the reviewer should verify that the procedure guidance ensures that any fluids, such 

as shield or borated water, fill their respective cavities according to design specifications. 

Additionally, for casks that require borated water to maintain subcriticality, the reviewer should 

ensure that the procedure guidance includes verification that the water to be used for cask 

reflood meets the minimum soluble boron content required by the Technical Specifications. 

Damaged Fuel (LOW Priority) 

The SAR should include appropriate additional measures for the potential presence of damaged 

fuel. Procedures should be designed to maximize worker protection from unanticipated 

radiation exposures or contaminants due to damaged fuel in accordance with ALARA principles 

and, to the maximum extent possible, prevent any uncontrolled releases to the environment. 

The following points outline the relevant safety concerns and an acceptable approach to 

address damaged fuel contingencies in cask unloading: 

 • The procedure descriptions should provide for fuel unloading under normal 

conditions. 

 • The unloading process should ensure that the fuel can be safely unloaded with 

regard to structural, criticality, thermal, and radiation protection considerations. 

This includes the provision for safe maintenance of the fuel and cask while any 

additional measures needed to address suspected damaged fuel are planned 

and implemented. 

 • The unloading process should reflect the potential for damaged fuel and 

changing radiological conditions. 
 9-13 

 • The process should include measures to check for and detect damaged fuel 

conditions (such as atmosphere samples) before opening the cask. (Note that 

fuel oxidation resulting from exposure to air at temperatures typical for dry 

storage is a known form of fuel degradation. Therefore, the presence of air in a 

cask designed to maintain an inert atmosphere indicates that the fuel may be 

degraded. The detection of fission gases is another indicator that the fuel may 

be degraded.) 

The process may establish sample result thresholds above which damaged fuel is suspected. 

Other technically sound methods may be used to check for potential air leakage paths. Such 

methods may include designs that monitor cask internal pressure or seal integrity and alert the 

licensee to a problem before oxidation could occur. However, this method may not address 

detection of potential fuel degradation resulting from other mechanisms (such as a cask drop 

accident). 

 • If the sample indicates normal conditions, the normal unloading process should 

be followed. 

 • If damaged fuel is suspected or found, the procedure description should stipulate 

that additional measures, appropriate for the specific conditions that include the 

canning of the damaged fuel, are to be planned, reviewed, and approved by the 

designated approval authority and implemented to minimize exposures to 

workers and radiological releases to the environment. These additional 

measures may include provision of filters, respiratory protection, and other 

methods to control releases and exposures in accordance with ALARA. 

Cooling, Venting, and Reflooding (LOW Priority) 

The reviewer should verify that the SAR describes applicable operational measures to control 

cask cooling, venting, and reflooding (when appropriate). Also, the reviewer should verify that 

these measures are consistent with the results of the structural, materials, and thermal 

evaluations in the SAR, respectively. Cask cooling, venting, and reflooding should not result in 

damage to the fuel. Operational measures may include external cooling of the confinement 

cask for initial temperature reduction, restricting reflood flow rates to control and limit internal 

cask pressure from steam formation, and limiting cooldown rates. 

Special attention should be devoted to reviews in this area since analysis of existing designs 

have predicted fuel temperatures during storage and transfer in excess of 533.15°K (500°F) for 

design-basis heat loads. Operational controls may be required to address the following 

potential effects during a cooldown and reflood evolution: 

 • Cask pressurization may occur as a result of steam formation as reflood water 

contacts hot surfaces. 

 • Excessive cooling rates may cause fuel cladding and fuel rod component 

damage and release of radioactive material as a result of stress (thermal, internal 

pressure, etc.) beyond material strengths (see SRP Section 8.4.17.1, “Cladding 

Temperature Limits”). 
 9-14 

 • Excessive cooling rates may induce thermal stress that causes gross 

deformation of the fuel assembly components and subsequent binding with the 

basket. 

 • Cask supply and vent line failures from inadequate design for pressure and 

temperature could result in radiological exposures and personnel hazards (e.g., 

steam burns). 

Fuel Crud (LOW Priority) 

The reviewer should verify that the procedure descriptions include contingencies for protection 

from fuel crud particulate material. Appendix E of ANSI/ANS 57.9 provides a short discussion of 

crud with respect to dry transfer systems. The unloading procedures should alert cask users to 

wait until any loose particles have settled and to slowly move the fuel assemblies to minimize 

crud dispersion in the SNF pool. Experience with wet unloading of boiling-water reactor (BWR) 

fuel after transportation has involved handling significant amounts of crud. This fine crud, which 

includes 60Co and 55Fe, will remain suspended in water or air for extended periods. The dry 

cask reflood process, during unloading of BWR fuel, has the potential to disperse crud into the 

fuel transfer pool and the pool area atmosphere, thereby creating airborne exposure and 

personnel contamination hazards. By contrast, no significant crud dispersal problems have 

been observed in handling pressurized-water reactor (PWR) fuel due to differences in the 

characteristics of crud on this type of fuel. 

ALARA (LOW Priority) 

The reviewer should verify that the procedure descriptions incorporate ALARA principles and 

practices. These may include provisions to perform radiological surveys, exposure and 

contamination control measures, temporary shielding, and suggested caution statements 

related to specific actions that could change radiological conditions. The reviewer should verify 

that any recommended surveys incorporate the applicable operating controls and limits 

described in Chapter 13, “Technical Specifications and Operation Controls and Limits 

Evaluation” of the SAR. 

Offsite Release (LOW Priority) 

Where applicable, the reviewer should verify that the SAR describes methods such as filtered 

ventilation, decontamination, or temporary containments to minimize offsite releases. The 

procedures should also provide for minimizing generation of radioactive waste. 

Administrative Programs (HIGH Priority) 

The applicant may request that the NRC approve an administrative program for cask unloading. 

NUREG-1745 provides requirements for such an administrative program. The reviewer should 

verify the proposed administrative program meets the requirements summarized in 

Section 9.5.1 of this SRP. 

9.6 Evaluation Findings 

The reviewer should examine the 10 CFR Part 72 acceptance criteria and provide a summary 

statement for each. These statements should be similar to the following model, as applicable: 
 9-15 

 F9.1 The [cask designation] is compatible with [wet/dry] loading and unloading. 

General procedure descriptions for these operations are summarized in 

Chapter(s) of the applicant’s safety analysis report (SAR). Detailed 

procedures will need to be developed and evaluated on a site-specific basis. 

 F9.2 The [welded/bolted lids or other features] of the cask allow ready retrieval of the 

spent fuel for further processing or disposal as required. 

 F9.3 The smooth surface [or other feature] of the cask is designed to facilitate 

decontamination. Only routine decontamination will be necessary after the cask 

is removed from the spent fuel pool. 

 F9.4 No significant radioactive waste is generated during operations associated with 

the independent spent fuel storage installation (ISFSI). Contaminated water from 

the spent fuel pool will be governed by the 10 CFR Part 50 license conditions. 

 F9.5 No significant radioactive effluents are produced during storage. Any radioactive 

effluents generated during the cask loading will be governed by the 10 CFR 

Part 50 license conditions. 

 F9.6 The content of the general operating procedures described in the SAR are 

adequate to protect health and minimize damage to life and property. Detailed 

procedures will need to be developed and approved on a site-specific basis. 

 F9.7 The radiation protection chapter of this SER assesses the operational restrictions 

to meet the limits of 10 CFR Part 20. Additional site-specific restrictions may 

also be established by the site licensee. 

The reviewer should provide a summary statement similar to the following: 

“The staff concludes that the generic procedures and guidance for the operation of the 

[cask designation] are in compliance with 10 CFR Part 72 and that the applicable 

acceptance criteria have been satisfied. The evaluation of the operating procedure 

descriptions provided in the SAR offers reasonable assurance that the cask will enable 

safe storage of spent fuel. This finding is based on a review that considered the 

regulations, appropriate regulatory guides, applicable codes and standards, and 

accepted practices.” 
 10-1 

10 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM EVALUATION 

10.1 Review Objective 

The acceptance tests and maintenance program review ensures that the applicant’s Safety 

Analysis Report (SAR) includes the appropriate acceptance tests and maintenance programs 

for the system. A clear, specific listing of these commitments will help avoid ambiguities 

concerning design, fabrication, and operational testing requirements when the U.S. Nuclear 

Regulatory Commission (NRC) staff conducts subsequent inspections. Acceptance tests may 

also be described in the applicable chapter of this Standard Review Plan (SRP). 

10.2 Areas of Review 

This chapter of the dry storage system (DSS) SRP provides guidance for use in evaluating the 

acceptance tests and maintenance programs outlined in the SAR. The acceptance tests 

demonstrate that the cask has been fabricated in accordance with the design criteria and that 

the initial operation of the cask complies with regulatory requirements. The maintenance 

program describes actions that the licensee needs to implement during the storage period to 

ensure that the cask performs its intended functions. 

As defined in Section 10.5, “Review Procedures,” a comprehensive evaluation may encompass 

the following acceptance tests and maintenance programs: 

Acceptance Tests

 Structural/Pressure Tests 

 Leak Tests 

 Visual and Nondestructive Examination Inspections 

 Shielding Tests 

 Neutron Absorber Tests 

 Thermal Tests 

 Cask Identification 

Maintenance Program

Inspection 

 Tests 

 Repair, Replacement, and Maintenance 

10.3 Regulatory Requirements 

This section presents a summary matrix of the portions of U.S. Code of Federal Regulations 

(CFR), Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel 

High-Level Radioactive Waste and Reactor-Related Greater Than Class C Waste,” Title 10, 

“Energy” (10 CFR Part 72) that are relevant to the review areas addressed by this chapter. The 

NRC staff reviewer should read the exact referenced regulatory language. Table 10-1 matches 

the relevant regulatory requirements associated with this chapter to the areas of review 

identified in the previous section. 
 10-2 

Table 10-1 Relationship of Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Regulations

72.82 

(d) 

72.122 

(a), (f) 

72.124 

(b) 72.162 72.212 (b)(8) 

72.232 

(b) 

72.236 

(c) 

72.236 

(g), (j), 

(k), (l) 

Acceptance Tests ● ● ● ● ● ●

Maintenance Program ● ● ●

Design Verification ● ● ● ● ● ●

10.4 Acceptance Criteria 

In general, the acceptance tests and maintenance programs outlined in the SAR should cite 

appropriate authoritative codes and standards. The staff has previously accepted the following 

as the regulatory basis for the design, fabrication, inspection, and testing of DSS components: 

System/Component Acceptable Regulatory Basis*

Confinement System • American Society of Mechanical Engineers (ASME), “Boiler 

and Pressure Vessel (B&PV) Code,” Section III, Division 1, 

2007 

• “American National Standard for Radioactive Materials – 

Leakage Tests on Packages for Shipment” (ANSI N14.5) 

Confinement Internals 

(e.g. basket) 

• ASME B&PV Code, Section III, Subsection NG 

Metal Cask Overpack • ASME B&PV Code, Section VIII 

Concrete Cask Overpack • American Concrete Institute (ACI), “Code Requirements for 

Structural Concrete” (ACI-318), “Code Requirements for 

Nuclear Safety Related Concrete” (ACI-349), as appropriate 

Other Metal Structures • ASME B&PV Code, Section III, Subsection NF 

• American Institute of Steel Construction (AISC), “Manual of 

Steel Construction” 

* The SAR should clearly identify any exceptions to the listed codes and standards (see SRP Chapter 13, 

“Technical Specifications and Operating Controls and Limits Evaluation”).

10.5 Review Procedures 

Introduction

Figures 10-1 and 10-2 present an overview of the evaluation process and can be used as a 

guide to assist in coordinating with the review disciplines. 
 10-3 

Figure 10-1 Overview of Acceptance Test Review Evaluation 

Chapter 1 – General 

Information Evaluation

• Codes and 

Standards

• Dimensions and 

Tolerances

• Component Materials

• Contents

Chapter 14 – Quality 

Assurance Evaluation

• Test Controls

Neutron Absorber Tests

Thermal Tests

Cask Identification

Structural/Pressure Tests

Leak Tests

Visual and Nondestructive Examination Inspections

Shielding Tests

Chapter 10.5.1 – Acceptance Tests Review Evaluation

Neutron Absorber Tests

Thermal Tests

Cask Identification

Structural/Pressure Tests

Leak Tests

Visual and Nondestructive Examination Inspections

Shielding Tests

Chapter 10.5.1 – Acceptance Tests Review Evaluation

Chapter 2 – Principal 

Design Criteria 

Evaluation

• Inspections and 

Documentation

• Test Procedures

• Codes and 

Standards

Chapter 5 –

Confinement 

Evaluation

• Fabrication 

Verification Leakage 

Rate

• Periodic Verification 

Leakage Rate

Chapter 6 – Shielding 

Evaluation

• Shielding Material 

Specifications

• Shielding Tests

Chapter 7 – Criticality 

Evaluation

• Dimensions and 

Tolerances

• Neutron Poisons

Chapter 3 – Structural 

Evaluation

• Codes and 

Standards

• Structural and 

Pressure Test

• Structural 

Components Tests

Chapter 4 – Thermal 

Evaluation

• Temperature

• Pressures

• Heat Transfer 

Features

• Thermal 

Performance 

Evaluations

Chapter 8 – Materials 

Evaluation

• Design Basis for 

SSCs Important to 

Safety

• Design Criteria for 

Safety Protection 

Systems

• Material Compatibility
 10-4 

Figure 10-2 Overview of Maintenance Program Review Evaluation 

Chapter 1 – General 

Information Evaluation

• Codes and 

Standards

• Dimensions and 

Tolerances

• Component Materials

• Contents

Inspection

Tests

Repair, Replacement, and Maintenance

Chapter 10.5.2 – Maintenance Program Review Evaluation

Chapter 2 – Principal 

Design Criteria 

Evaluation

• Surveillance

Chapter 6 –

Shielding 

Evaluation

• Shielding Tests

• Shielding Material 

Specifications

Chapter 7 –

Criticality 

Evaluation

• Dimensions and 

Tolerances

• Neutron Poisons

Chapter 3 – Structural 

Evaluation

• Codes and 

Standards

• Structural and 

Pressure Test

• Structural 

Components Tests

Chapter 4 – Thermal 

Evaluation

• Temperature

• Pressures

• Heat Transfer 

Features

Chapter 8 –

Materials 

Evaluation

• Design Basis for 

SSCs Important 

to Safety

• Design Criteria for 

Safety Protection 

Systems

• Material 

Compatibility

Chapter 5 –

Confinement 

Evaluation

• Fabrication 

Verification 

Leakage Rate

• Periodic 

Verification 

Leakage Rate

Chapter 9 –

Confinement 

Evaluation

• Periodic Testing

• Replacement 

Component 

Testing
 10-5 

The review procedures described in this section are presented in a format intended to facilitate 

a single, independent review. Although one or more individual(s) may be tasked with preparing 

the corresponding section of the safety evaluation report (SER) related to the proposed 

acceptance tests and maintenance program, all review team members should examine the 

related information presented in the SAR. Information in the SAR related to the acceptance 

tests may be located in the chapters related to specific disciplines (e.g. SAR Chapter 4, 

“Thermal Evaluation”) and/or in SAR Chapter 10, “Acceptance Tests and Maintenance 

Program.” Reviewers should devote special attention to those tests (or the lack of tests) that 

affect their functional area of review. If the descriptions included in the SAR are not sufficiently 

detailed to allow a complete evaluation concerning fulfillment of the acceptance criteria, 

reviewers should request additional information from the applicant. 

In general, applicants commit to design, construct, and test the system under review to the 

codes and standards identified in SAR Chapter 2, “Principal Design Criteria.” The NRC does 

not generally review specific test and maintenance procedures as part of the licensing process; 

however, the applicant is expected to describe (in the SAR) certain elements of the proposed 

test and maintenance programs. The staff may inspect selected portions of test procedures as 

part of its onsite activities. 

The following subsections provide representative examples of test and maintenance program 

elements that should be subject to licensing review. If included in the SAR, each of these tests 

and maintenance elements should be reviewed to ensure that the applicant has identified the 

purpose of the test, explained the proposed test method (including any applicable standard to 

which the test will be performed), defined the acceptance criteria and bases for the test, and 

described the actions to be taken if the acceptance criteria are not satisfied. 

10.5.1 Acceptance Tests (Priority – as indicated) 

The following guidance is presented on the basis of tests deemed acceptable by the staff in 

previous SAR reviews. The guidance is based on operational experience and the knowledge 

from past licensing reviews. Alternative tests and criteria may be used if the SAR provides 

appropriate explanation and adequate justification. Additional tests and criteria may be needed, 

depending on the operational experience and uniqueness of the design proposal. 

10.5.1.1 Structural/Pressure Tests 

(MEDIUM Priority) Lifting trunnions should be fabricated and tested in accordance with ANSI 

N14.6, “American National Standard for Radioactive Materials-Special Lifting Devices for 

Shipping Containers Weighing 10,000 pounds (4,500 Kilograms) or More.” Site-specific details 

of the pool and lifting procedures may enable the cask to be considered a non-critical load, as 

defined by this standard. Generally, however, the cask is considered a critical load during its 

handling in the pool. Consequently, trunnion testing should be performed at a minimum of 

150 percent of the maximum service load, if redundant lifting is employed or 300 percent of the 

service load if non-redundant lifting applies. These load tests should be performed to ensure 

that the trunnions and cask are conservatively constructed and provide an adequate margin of 

safety when filled with SNF. Trunnion load testing should also be performed annually for the 

transfer cask and at least one year before use for the storage cask. Load testing of integral 

trunnions is not required once the loaded storage cask has been placed on the pad. 

Restrictions on cask lifting resulting from these tests should be included in Chapter 13, 

“Technical Specifications and Operating Controls and Limits Evaluation,” of the SAR and the 

related SER prepared by the NRC staff. SAR Chapter 10, “Acceptance Tests and Maintenance 
 10-6 

Program Evaluation” should explicitly state the testing values. Periodical NDE, in lieu of annual 

load tests, is acceptable for the trunnion provided that other conditions as specified in ANSI 

N14.6 are also met. 

(MEDIUM Priority) The entire confinement boundary should be pressure tested hydrostatically 

or pneumatically to 125 or 110 percent of the design pressure, respectively. The pressure test 

should be performed in accordance with governing code associated with the confinement 

boundary, which typically has been ASME B&PV Code, Section III, Division 1, Subsection NB or 

NC. The test pressure should be maintained for a minimum of 10 minutes, after which a visual 

inspection should be performed to detect any leakage. SAR Chapter 10, “Acceptance tests and 

Maintenance Program Evaluation” should clearly specify the hydrostatic and pneumatic test 

pressures. The helium leakage test, per ANSI 14.5 is not considered as a substitute for the 

Code required pressure test, and conversely, the Code pressure test is not a substitute for the 

helium leakage test. If a shop pressure test isn’t performed and only a field pressure test is 

performed after the first closure weld is made, the staff has accepted the shop helium leakage 

test as meeting the pressure test acceptance criteria of no leakage for the shell welds since they 

are generally inaccessible in the field. 

(LOW Priority) Some casks contain a neutron shielding material that may off-gas at higher 

temperatures. Such material is usually contained inside a thin steel shell to prevent loss of 

mass and provide protection from minor accidents and natural phenomenon events. Rupture 

disks or relief valves are generally provided to prevent catastrophic failure of this shell. The 

shell should be tested to 125 percent of the rupture disk burst pressure, which is usually 

equivalent to 125 percent of the shell design pressure. The SAR should clearly specify the 

burst pressure for the rupture disk, along with its coincident burst temperature and tolerance on 

burst pressure. 

(HIGH Priority) Some cask designs use ferritic steels that are subject to brittle fracture failures at 

low temperature. ASME B&PV Code, Section II, Part A, contains procedures for testing ferritic 

steel used in low temperature applications. NUREG/CR-1815, “Recommendations for 

Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Up to Four 

Inches Thick,” provides staff guidance concerning materials and thickness ranges subject to 

brittle fracture testing. On the basis of guidance in NUREG/CR-1815, Section 5.1.1, the NRC 

established two methods for identifying suitable materials: 

 • The nil-ductility transition (NDT) temperature must be determined by either direct 

measurement, (American Society for Testing and Materials’ (ASTM) “Method of 

Conducting Drop Weight Test to Determine Nil-ductility Transition Temperature 

for Ferritic Steel” [ASTM E-208]) or indirect measurement (“Dynamic Tear 

Testing of Metallic Materials” [ASTM E-604]), and the minimum operating 

temperature of the steel must be specified as 28°C (50°F) higher than the NDT. 

 • The NRC staff accepts ASME Charpy testing procedures for verification of the 

material’s minimum absorbed energy. Acceptable energy absorption values and 

test temperatures of Charpy, V-Notch impact tests are listed in the ASME B&PV 

Code, Section II, SA-20, “Specifications for General Requirements for Steel 

Plates for Pressure Vessels” Table A1.15. Coordinate with the thermal review 

(Chapter 4 of this SRP) to ensure that the applicant selected the correct 

temperatures for the tests and that the SAR specifies the method of testing. 
 10-7 

For casks with ferritic steel walls thicker than 102 mm (4 in.), the guidance provided in 

NUREG/CR-3826, “Recommendations for Protecting Against Failure by Brittle Fracture in 

Ferritic Steel Shipping Containers Greater than Four Inches Thick,” should be followed. 

10.5.1.2 Leak Tests (MEDIUM Priority) 

The licensee should perform leak tests on all confinement boundaries except as excluded in 

Chapter 8, “Materials Evaluation” - Section 8.9.2, which only applies to the closure welds 

typically made in the field. For all-welded cask confinements, the NRC staff has, with adequate 

justification, considered it acceptable for licensees to omit leak testing of the second cask 

closure weld and the seal welds for the closure plates of the purge and vent valves (if not 

potentially pressurized at the time of welding). For such cases, leak testing must show that the 

inner closure weld meets the leakage limits. A fabrication leak test should be performed on 

every canister in the shop to ensure that the tested leakage rate is compatible with the 

regulatory dose limits at the controlled area boundary, 10 CFR 72.236(d), (i), and (j). 

Leakage criteria in units of Pa.m3

/s or reference cm3/s must be at least as restrictive as those 

specified in the principal design criteria (in SAR Chapter 2). The SAR should also indicate the 

general testing methods (e.g., pressure increase, mass spectrometer) and required sensitivities. 

If cask closure depends on more than one seal (e.g., lid, vent port, drain port), the leakage 

criteria should ensure that the total leakage is within the design requirements. Leak testing 

should be conducted in accordance with ANSI N14.5. 

10.5.1.3 Visual and Nondestructive Examination Inspections 

(HIGH Priority) Reviewers should verify the applicant’s commitment to fabricate and examine 

cask components in accordance with an accepted design standard such as ASME B&PV Code, 

Section III or VIII. These sections define the examination requirements mentioned in Section II, 

“Materials Specifications and Properties”; Section V, “NDE Specifications and Procedures”; and 

Section IX, “Qualification Standard for Welding and Brazing Procedures, Welders, Brazers, and 

Welding and Brazing Operators.” The following guidance assumes that the ASME B&PV Code 

is applicable to the cask being reviewed. 

(HIGH Priority) The nondestructive examination (NDE) of weldments must be well-characterized 

on drawings, using standard NDE symbols and/or notations (see American Welding Society’s 

(AWS) “Standard Symbols for Welding, Brazing, and Nondestructive Examination” [AWS A2.4]). 

Each fabricator should be required to establish and document a detailed, written weld inspection 

plan in accordance with an approved quality assurance (QA) program that complies with 

10 CFR Part 72, Subpart G. The inspection plan should include visual (VT), dye penetrant (PT), 

magnetic particle (MT), ultrasonic (UT), and radiographic (RT) examinations, as applicable. The 

inspection plan should identify welds to be examined, the examination sequence, type of 

examination, and the appropriate acceptance criteria as defined by either the ASME B&PV 

Code, or an alternative approach proposed and justified by the applicant. Inspection personnel 

should be qualified, in accordance with the current revision of the American Society for 

Nondestructive Testing’s (SNT) “Personnel Qualification and Certification in Nondestructive 

Testing” (SNT-TC-1A), as specified by the ASME B&PV Code. All weld-related NDE should be 

performed in accordance with written and approved procedures. Fabrication controls and 

specifications should be in-place and field tested to prevent post-welding operations (such as 

grinding) from compromising the design requirements (such as wall thickness). 
 10-8 

(HIGH Priority) Confinement boundary non-closure welds should meet the requirements of 

ASME B&PV Code, Section III, Division 1, Subsections NB or NC, Article NB/NC-5200, 

“Required Examination of Welds for Fabrication and Preservice Baseline.” This section requires 

volumetric examination and either PT or MT for all Category A and most Category B or 

Category C welded joints in vessels, and longitudinal or full penetration welded joints in other 

components. The ASME-approved specifications for RT, UT, PT, and MT are detailed in ASME 

B&PV Code, Section V, Articles 2, 4, 6, and 7, respectively. 

(HIGH Priority) Acceptance standards for nondestructive testing should be in accordance with 

ASME B&PV Code, Section III, Division 1, Subsection NB or NC -5300. Testers should reject 

unacceptable imperfections (such as a crack, a zone of incomplete fusion or penetration, 

elongated indications with lengths greater than specified limits, and rounded indications in 

excess of the limits in ASME B&PV Code, Section III, Division 1, Appendix VI). Repaired welds 

should be reexamined in accordance with the original examination method and associated 

acceptance criteria. 

(HIGH Priority) For confinement welds that cannot be volumetrically examined using RT, the 

licensee may use 100 percent UT. The ASME-approved UT specifications are detailed in 

ASME B&PV Code, Section V, Article 4. Acceptance criteria should be defined in accordance 

with ASME B&PV Code, Section III, Division 1, Subsection NB or NC-5330, “Ultrasonic 

Acceptance Standards.” Cracks, lack of fusion, or incomplete penetration are unacceptable, 

regardless of length. 

(HIGH Priority) The NRC has accepted multiple surface examinations of welds, combined with 

helium leak tests for inspecting the final redundant seal welded closures. 

(HIGH Priority) For confinement internals, the licensee should perform all NDE testing in 

accordance with ASME B&PV Code, Section III, Division 1, Subsection NG. 

(LOW Priority) Nonconfinement welds (which exclude welds of confinement internals) should 

meet the requirements of ASME B&PV Code, Section III, Subsection NF, or Section VIII, 

Division 1, as applicable. The required volumetric examination of welds is either RT or UT, as 

discussed in ASME B&PV Code, Section III, NF-5200, and Section VIII, UW-11. The 

appropriate specifications from ASME B&PV Code, Section V, are invoked in Article 2 for RT 

and in Article 5 for UT. Acceptance standards for RT are detailed in ASME B&PV Code, 

Section III, Subsection NF, NF-5320, “Radiographic Acceptance Standards,” and for UT in 

NF-5330, “Ultrasonic Acceptance Standards.” For Section VIII weldments, RT acceptance 

criteria should be in accordance with ASME B&PV Code, Section VIII, Division 1, UW-51, and 

the repair of unacceptable defects should be in accordance with UW-38. Repaired welds 

should be reexamined in accordance with the original acceptance criteria. 

(LOW Priority) Nonconfinement welds that cannot be examined using RT should undergo UT in 

accordance with ASME B&PV Code, Section V, Article 4. Acceptance criteria should be in 

accordance with ASME B&PV Code, Section VIII, Division 1, UW-53 and Appendix 12, and the 

repair of unacceptable defects should be in accordance with UW-38. Repaired welds should be 

reexamined in accordance with the original examination methods and associated acceptance 

criteria. If applicable, the SAR should also justify the rationale for not requiring RT examination 

of these welds. 

(LOW Priority) Nonconfinement welds for cask system components that are designed and 

fabricated in accordance with ASME B&PV Code, Section III, that cannot be examined using RT 
 10-9 

or UT should undergo PT or MT examination in accordance with ASME B&PV Code, Section V, 

Articles 6 and 7, respectively. Acceptance criteria should be in accordance with Articles 

NF-5350 and NF-5340, respectively. Repaired welds should be reexamined in accordance with 

the original acceptance criteria. If applicable, the SAR should also justify the rationale for not 

requiring volumetric inspection techniques (RT or UT) for these welds. 

(Low Priority) Nonconfinement welds may also be welded, repaired and examined in 

accordance with AWS D1.1, Structural Welding Code – Steel, D1.3, Structural Welding Code – 

Sheet Steel and D1.6, Structural Welding Code – Stainless Steel. Use of these standards shall 

be called out on the licensing drawings. 

(LOW Priority) Finished surfaces of the cask should be visually examined in accordance with 

the ASME B&PV Code Section V, Article 9. For welds examined using VT, the acceptance 

criteria should be in accordance with ASME B&PV Code, Section VIII, Division 1, UW-35 and 

UW-36, or NF-5360, “Acceptance Standards for Visual Examination of Welds.” 

(HIGH for confinement/LOW for non-confinement) The licensee should use PT to detect 

discontinuities (such as cracks, seams, laps, laminations, and porosity) that open to the surface 

of nonporous metals. PT should be performed in accordance with ASME B&PV Code, 

Section V, Article 6. Acceptance criteria for PT examination of confinement welds should be in 

accordance with ASME B&PV Code, Section III, Subsection NB/NC, Article NB/NC-5350. 

Repair procedures should be in accordance with NB/NC-4450 of the ASME B&PV Code, 

Section III. Acceptance criteria for PT examination of nonconfinement welds should be in 

accordance with ASME B&PV Code, Section VIII, Division 1, Appendix 8, or NF-5350, “Liquid 

Penetrant Acceptance Standards.” Repair procedures should be in accordance with ASME 

B&PV Code, Section III or NF-2500, “Examination and Repair of Material,” and NF-4450, 

“Repair of Weld Material Defects.” 

10.5.1.4 Shielding Tests (LOW Priority) 

The materials that comprise the DSS should sufficiently maintain their physical and mechanical 

properties during all conditions of operations. DSS gamma shielding materials (e.g., lead) 

should not experience slumping or loss of shielding effectiveness to an extent that compromises 

safety. The shield should perform its intended function throughout the licensed service period. 

DSS materials used for neutron shielding should be designed to perform their safety function 

without degradation, gas release, or physical alteration for the full term of the license. Tests are 

required to ensure these conditions are met. 

Tests of the effectiveness of both the gamma and neutron shielding may be required if, for 

example, the cask contains a poured lead shield or a special neutron absorbing material. In 

such instances, the SAR should describe any scanning or probing with an auxiliary source for 

the purpose of characterizing the shielding. This shield testing should be done for every cask 

that uses poured shielding material, to demonstrate proper fabrication in accordance with the 

design drawings. Alternatively, the applicant may propose an alternate testing program for 

fabricated casks with appropriate justification. 

Dose measurements of loaded SNF, in lieu of an auxiliary source, may be used to verify 

shielding effectiveness with appropriate scanning of the shield and appropriate testing program 

that considers the actual source strength of the loaded contents.. 
 10-10 

10.5.1.5 Neutron Absorber Tests (HIGH Priority) 

Neutron absorber materials require both qualification and acceptance testing to provide 

assurance that the control of criticality by absorbing thermal neutrons will be assured in systems 

designed for nuclear fuel storage, transportation or both. Both qualification and acceptance 

testing are in general as described in ASTM Designation C1671, “Standard Practice for 

Qualification and Acceptance of Boron Based Metallic Neutron Absorbers for Nuclear Criticality 

Control for Dry Storage Systems and Transportation Packaging.” 

Acceptance tests are used to ensure that material properties for plates and other shapes 

produced in a given production run are in compliance with the materials requirements of the 

application. In one sense, acceptance tests verify that the material of a given production run 

has yielded products that have been shown to be like the products that were used in the 

qualification testing. Acceptance tests are used to ensure that the production process is 

operating in a satisfactory manner, and they use statistical data for selected measurable 

parameters. For all boron-containing absorber materials, acceptance tests should (a) verify 10B 

content and uniformity, (b) require visual examinations to establish only acceptable levels of 

defects are present from cracks, porosity, blisters, or foreign inclusions, and (c) make 

dimensional (e.g., plate thickness which is important to the areal density). 

Some materials may obtain 100 percent credit for the amount of 10B that is shown to be present 

in the absorber materials. This level of credit is sometimes called 90 percent credit because the 

credit level refers to a manner in which K-effective calculations are conducted and in these 

calculations, any absorber is given a 10 percent penalty before being used in the calculation. 

Likewise other materials that are given only 82 percent credit are called materials with 

75 percent credit. For purposes of obtaining high levels (100 percent) of credit, the amount of 

10B, which is the absorber species, is assessed in boron-containing absorber materials using 

neutron attenuation testing. 

Neutron attenuation tests are calibrated using appropriate standards such as those based on 

(coated with) zirconium diboride (ZrB2) plates to ensure the accuracy of the measured values. 

Approved substitutes may be used for the attenuation tests. These include tests such as 

chemical analysis, provided that (1) both the neutron attenuation tests and the alternative tests 

have at least the sensitivity of tests specified in C-1671 and (2) the alternate form of testing is 

regularly bench marked against calibrated neutron attenuation tests. Chemical analyses should 

also include spectrochemical analysis for material impurity levels and 10B content. Uniformity is 

assessed using statistical sampling techniques that ensure that the entire plate of material and 

all plates in a lot meet a 95/95 criterion, which means that a test result has a 95 percent 

likelihood of containing the minimum required amount of 10B and that this is known at the 95 

percent confidence level. 

The reviewer should confirm that the calculation of minimum poison content (e.g., poison areal 

density) conservatively accounts for tolerance limits on material density, poison concentration, 

and component dimensions. It is noted that thickness tolerances on rolled plates, sheets or 

shape are typically on the order of ± 10 percent. The acceptance testing should provide a 

representative sampling of coupons for plates or sheets from a given lot. Statistical sampling 

can be used to the extent practical, using test locations on a coupon that will account for local 

variations or anomalies within the coupon and hence within the plates represented by the 

coupon. Adequate numbers of samples should be taken to ensure the confidence level required 

for the application. 
 10-11 

Acceptance Testing of Fabricated Materials for 75-Percent Boron Credit

For multi-phase absorber materials analyzed with 75-percent poison credit (or less) the reviewer 

should confirm that acceptance testing is consistent with the following: 

 • The effective 10B content should be verified from plate coupons by either 

(a) neutron attenuation testing, or (b) chemical assay to determine boron content 

with mass spectrometric analysis for isotopic composition. 

 • Sufficient coupons should be taken for acceptance testing to justify the level of 

credit given. Rejection of a coupon should result in rejection of the plate from 

which it is taken. Sampling may be reduced to lesser percentages of the 

coupons taken (e.g., to 50 percent of all coupons) after acceptance of all 

coupons in the first 25 percent of the lot. A rejection during reduced inspection 

should invoke a 100 percent inspection for coupons from that lot. 

 • A visual examination of all plates for defects should be conducted. 

Acceptance Testing for Greater Than 75 Percent Boron Credit

For acceptance testing of borated absorbers at levels of poison credit beyond 75 percent, the 

extent of the acceptance testing and inspection is enhanced. Some of the data helpful in 

meeting the guidance in C-1671 Sec 5.3.4 are as follows: 

 • The effective 10B content is verified by neutron attenuation testing of coupons. 

An adequate number of coupons should be acceptance tested for each lot of 

materials to statistically demonstrate that the 95/95 criterion is satisfied for the 

minimum required 10B content. The minimum areal density is specified in the 

SAR. Note that if the coupon from a plate fails the single neutron attenuation 

measurement, the associated plate is rejected unless acceptable alternative 

testing is done with acceptable results. 

 • Sufficient coupons should be taken to satisfy the 95/95 criterion. For example, 

coupons are taken from at least every other plate unless justification for fewer is 

given. Measurements are made on samples taken from 100 percent of all 

coupons. Rejection of a coupon should result in rejection of the plate. Sampling 

may be reduced to 50 percent of all coupons after acceptance of all coupons in 

the first 25 percent of the lot. A rejection during reduced inspection should 

invoke a return to 100 percent inspection for that lot. 

 • A statistical analysis of the neutron attenuation results should be performed by 

the applicant for all plates in a lot. This analysis shall show that the lot meets the 

95/95 criterion. That is, using a one-sided tolerance limit factor for a normal 

distribution with at least 95 percent probability, the areal density is greater than or 

equal to the specified minimum value with 95 percent confidence level. Failure to 

meet this acceptance criterion of this statistical analysis shall result in rejection of 

the entire lot for use at the 100 percent (90 percent credit in K-effective 

calculations). Applicants may choose to convert all areal densities determined by 

neutron attenuation to a volume density by dividing by the thickness of the 

coupon. The one side tolerance limit of volume density with 95 percent 

probability and 95 percent confidence may then be determined. The minimum 
 10-12 

specified value of the areal density may be divided by the 95/95 lower tolerance 

limit of 10B volume density to arrive at the minimum plate thickness. Hence, all 

plates which have any locations thinner than this minimum shall be rejected, and 

those equal to or thicker may be accepted. 

 • A visual examination of all plates for defects should be verified. 

The reviewer should refer to Section 8.4.13.2 of this SRP regarding how to compute per level of 

credit. 

10.5.1.6 Thermal Tests (LOW Priority) 

Depending on the details of the DSS design and the ability to determine its heat removal 

capability through thermal analysis, testing may be required to verify performance. The 

applicant should establish acceptance criteria on the basis of the conditions of the test (e.g., test 

heat loading, ambient conditions). SAR Chapter 4, “Thermal Evaluation,” should discuss the 

correlation between test performance and actual loading conditions to avoid ambiguous or 

unreviewed analysis after the test data are obtained. 

10.5.1.7 DSS Identification (LOW Priority) 

The vendor/licensee must mark the DSS with a model number, unique identification number, 

and empty weight. Generally this information will appear on a data plate, which should be 

detailed in one of the drawings included in SAR Chapter 1, “General Description.” In addition, 

the vendor/licensee should mark the exterior of shielding casks or other structures that may hold 

the confinement while it is in storage. This marking should provide a unique, permanent, and 

visible number to permit identification of the cask stored therein. 

10.5.2 Maintenance Program (LOW Priority) 

DSSs are typically designed as passive units requiring minimal maintenance. The SAR should 

address the following areas, as applicable: 

10.5.2.1 Inspection 

Usually, the DSS has at least one monitoring system (e.g., pressure, temperature, dosimetry). 

The SAR should discuss how such systems will be used to provide information regarding 

possible off-normal events and what surveillance actions may be necessary to ensure that these 

systems function properly. Detailed procedures will be developed and implemented by the 

licensee at the site. 

The SAR should describe routine periodic visual surface and weld inspections, which should be 

limited to the readily accessible surfaces (i.e., the exterior surface of the DSS and all surfaces of 

empty transfer casks). In addition, the SAR should discuss inspection of lifting and rotating 

trunnion load-bearing surfaces. 

10.5.2.2 Tests 

The SAR should describe any periodic tests of DSS components or calibration of monitoring 

instrumentation, as well as periodic tests to verify shielding, thermal, and confinement 

capabilities. The applicant should otherwise justify that aging and degradation of materials 
 10-13 

related to the shielding, confinement, and thermal designs are not credible during the licensed 

period of the DSS. The SAR should also describe procedures for any applicable periodic 

testing of neutron poison effectiveness. As an alternative to the licensee’s periodic testing of 

neutron poison effectiveness, the applicant may show continued poison effectiveness in the 

manner described in Section 7.5.3.2 of this SRP. The qualification tests of the poison material, 

discussed in Section 8.4.13.3 of this SRP, may also be useful in showing the material’s 

continued effectiveness. 

In addition, the SAR should discuss any routine testing of support systems (e.g., vacuum drying, 

helium backfill, and leak testing equipment). 

10.5.2.3 Repair, Replacement, and Maintenance 

The SAR should discuss the repair and replacement of DSS components, as may be required 

during the lifetime of the storage and transfer casks. This discussion should include methods of 

repair or replacement, testing procedures, and acceptance criteria. The SAR should also 

describe procedures for routine maintenance (such as lubrication and re-application of corrosion 

inhibiting materials in the event of scratches) through the expiration of the service life of the 

equipment. Such information is also often included in SAR Chapter 12, “Accident Analyses,” 

which describes actions to be taken following an off-normal event or accident-level condition. 

10.6 Evaluation Findings 

The 10 CFR Part 72 acceptance criteria should be reviewed with a summary statement 

provided for each. These statements should be similar to the following model, as applicable: 

 F10.1 Chapter(s) of the SAR describe(s) the applicant’s proposed program for 

preoperational testing and initial operations of the (DSS designation). 

Chapter(s) discuss the proposed maintenance program. 

 F10.2 Structures, systems, and components (SSCs) important to safety will be 

designed, fabricated, erected, tested, and maintained to quality standards 

commensurate with the importance to safety of the function they are intended to 

perform. Chapter of the SAR identifies the safety importance of SSCs, and 

Chapter(s) present(s) the applicable standards for their design, 

fabrication, and testing. 

 F10.3 The applicant/licensee will examine and/or test the (DSS designation) to ensure 

that it does not exhibit any defects that could significantly reduce its confinement 

effectiveness. Chapter(s) of the SAR describe(s) this inspection and 

testing. 

 F10.4 The applicant/licensee will mark the DSS with a data plate indicating its model 

number, unique identification number, and empty weight. Drawing in 

SAR Chapter illustrates and/or describes this data plate. 

The reviewer should provide a summary statement similar to the following: 

“The staff concludes that the acceptance tests and maintenance program for the (DSS 

designation) are in compliance with 10 CFR Part 72 and that the applicable acceptance 

criteria have been satisfied. The evaluation of the acceptance tests and maintenance 
 10-14 

program provides reasonable assurance that the cask will allow safe storage of spent 

fuel throughout its licensed or certified term. This finding is reached on the basis of a 

review that considered the regulation itself, appropriate regulatory guides, applicable 

codes and standards, and accepted practices.” 
 11-1 

11 RADIATION PROTECTION EVALUATION 

11.1 Review Objective 

This chapter describes the radiation protection evaluation requirements and considerations of 

the proposed dry storage system (DSS). As used here, radiation protection refers to 

organizational, design, and operational elements that are primarily intended to limit radiation 

exposures from normal operations and anticipated occurrences. The evaluation of the 

radiological consequences for accidents is addressed in Chapter 12, “Accident Analyses 

Evaluation” of this SRP. 

The primary objectives of the radiation protection evaluation are to determine whether the 

design features and proposed operations meet the following criteria: 

 • the proposed DSS radiation protection features meet the U.S. Nuclear 

Regulatory Commission (NRC) design criteria for direct radiation; 

 • the applicant has proposed engineering features and operating procedures for 

the DSS that will ensure occupational exposures will remain ALARA; and 

 • the radiation doses to the general public will meet regulatory standards during 

both normal conditions and anticipated occurrences. 

In independent spent fuel storage installation (ISFSI) operations, the major mode of radiation 

exposure associated with spent nuclear fuel (SNF) storage cask handling is from direct 

radiation. Because of the cask design requirements, radionuclides are not expected to be 

released from the cask during either normal operations or design-basis accidents (DBAs). 

11.2 Areas of Review 

This chapter of the DSS Standard Review Plan (SRP) provides guidance for use in evaluating 

the radiation protection capabilities of the proposed cask system. The following outline shows 

the areas of review addressed in Section 11.4, “Acceptance Criteria,” and Section 11.5, “Review 

Procedures,” that may be encompassed in a comprehensive radiation protection review: 

Radiation Protection Design Criteria and Features

Occupational Exposures

Exposures at or Beyond the Controlled Area Boundary

 Normal Conditions 

 Accident Conditions and Natural Phenomenon Events 

ALARA

 Design Considerations 

 Engineering Controls and Procedures 
 11-2 

11.3 Regulatory Requirements 

This section presents a summary matrix of the portions of U.S. Code of Federal Regulations 

(CFR) Parts 20 and 72 that are relevant to the review areas addressed by this chapter. The 

NRC staff reviewer should read the exact referenced regulatory language. Virtually the entire 

contents of 10 CFR 20 “Standards for Protection Against Radiation” are also applicable to this 

review. Tables 11-1 and 11-2 match the relevant regulatory requirements associated with this 

chapter to the areas of review identified in the previous section. 

Table 11-1 Relationship of 10 CFR Part 20 Regulations and Areas of Review 

Areas of 

Review

10 CFR Part 20 Regulations

20.1101 20.1201 

(a) 20.1207 20.1208 

20.1301 

(a), (b), 

(d) 

20.1302 

(a) 20.1406 20.1501 

(a)(1) 20.1701 20.1702 

Radiation 

Protection 

Design 

Criteria and 

Features 

● ● ● ● ●

Occupational 

Exposures ● ● ● ● ● ●

Exposures at 

or Beyond 

the 

Controlled 

Area 

Boundary 

● ● ● ● 

ALARA ● ● ● ●

Table 11-2 Relationship of 10 CFR Part 72 Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Regulations

72.104 72.106(b) 72.126 

(a), (d) 72.236(d) 

Radiation Protection Design Criteria and 

Features ● ● ● ●

Occupational Exposures ● ● 

Exposures at or Beyond the Controlled 

Area Boundary ● ● ● ●

ALARA ● ● ●
 11-3 

11.4 Acceptance Criteria 

This section describes the acceptance criteria used for review of radiation protection features of 

and programs proposed for use with a DSS. These criteria are organized according to the 

areas of review specified in Section 11.2 of this chapter. The reviewer should note that some 

overlap exists between acceptance criteria for radiation protection and those related to Chapter 

5, “Confinement Evaluation,” and Chapter 6, “Shielding Evaluation,” of this SRP; therefore, the 

reviews of these chapters should be coordinated. 

11.4.1 Radiation Protection Design Criteria and Features 

Limitations on dose rates associated with direct radiation from the cask are established on the 

basis of the shielding and confinement evaluations to satisfy the regulatory requirements for 

dose limits to individuals located beyond the controlled area boundary (10 CFR 72.104). 

11.4.2 Occupational Exposures 

Estimated dose rates should be provided in Chapter 6, “Shielding Evaluation,” of the Safety 

Analysis Report (SAR) for representative points within the restricted area(s) (e.g. on and near 

the transfer cask and storage overpack surfaces) as well as at or beyond the perimeter of the 

controlled area. The radiation protection review includes a dose assessment that incorporates 

findings of the shielding review, as applicable. Individual and collective doses should be 

calculated. 

All individual doses to workers should be well below the dose limits specified in 

10 CFR 20.1201. Collective doses should be consistent with the objectives contained in a wellstructured ALARA program. The information provided by the applicant should allow for the 

determination of compliance with these criteria. To assess the applicant’s occupational dose 

calculations, the reviewer should check such things as the number of workers specified for a 

task and the time specified for performing the task being reasonable. 

11.4.3 Exposures at or Beyond the Controlled Area Boundary 

a. Normal Conditions: 

For normal operations and anticipated occurrences, the estimated dose to any real 

individual located at or beyond the controlled area boundary may not exceed the 

following values specified in 10 CFR 72.104(a): 

Whole body 0.25 mSv/yr (25 mrem/yr) 

Thyroid 0.75 mSv/yr (75 mrem/yr) 

Other organ 0.25 mSv/yr (25 mrem/yr) 

For purposes of the DSS review, the calculated doses must include both direct radiation 

and any planned discharges of radioactive material. 
 11-4 

b. Accident and Natural Phenomenon Events: 

Radiation shielding and confinement features should be provided sufficient to meet the 

requirements of 10 CFR 72.106(b). Any individual located on or beyond the nearest 

boundary of the controlled area may not receive the following dose from any DBA: 

The more limiting of 

TEDE or 

Sum of the DDE and the CDE to any individual 

organ or tissue (other than the lens of the eye) 

0.05 Sv (5 rem) 

0.5 Sv (50 rem) 

Lens of the eye 0.15 Sv (15 rem) 

Shallow Dose Equivalent (SDE) to skin or any 

extremity 

0.5 Sv (50 rem) 

11.4.4 ALARA 

For any new design or design change, the ALARA discussion should demonstrate how the 

design or design change 

 • accounted for radiation protection, technological, and economic considerations; 

and 

 • to the extent practicable, employed engineering controls and procedures that 

were founded upon sound radiation protection principles. 

11.5 Review Procedures 

The interrelationship of the radiation protection review with other disciplines is shown in 

Figure 11-1. 

11.5.1 Radiation Protection Design Criteria and Features for the Transfer Cask 

and Storage Cask (MEDIUM Priority) 

The reviewer should read the general description and functional features of the cask presented 

in Chapter 1, “General Description,” of the SAR. In addition, Chapter 2, “Principal Design 

Criteria,” of the applicant’s SAR should be reviewed as well as any additional detail regarding 

radiation protection provided in the Shielding and Confinement chapters of the SAR. If not 

previously discussed, the following additional criteria should be presented in Chapter 11, 

Radiation Protection, of the SAR. 
 11-5 

Figure 11-1 Overview of the Radiation Protection Evaluation

Chapter 1 – General 

Information Evaluation 

Chapter 9 – Operating 

Procedures Evaluation 

• Radiation Protection 

Features 

• Dose Assessment 

• Health Physics Program 

• Dose Rate Limits 

• ALARA Procedures 

• Dose Monitoring Program 

Chapter 10 – Acceptance 

Tests and Maintenance 

Program Evaluation 

• ALARA Concepts 

Chapter 2 – Principal 

Design Criteria Evaluation 

• ALARA Requirements 

Chapter 5 – Confinement 

Evaluation 

• Estimated Dose Rates 

• Features 

Chapter 6 – Shielding 

Evaluation 

• Estimated Dose Rates 

• Design Features 

ALARA

• Design Considerations 

• Engineering Controls and 

Procedures 

Exposures at or Beyond the 

Controlled Area Boundary 

• Normal Conditions 

• Accident Conditions and Natural 

Phenomenon Events 

Operational 

Exposures 

Radiation Protection 

Design Criteria and 

Features 

Chapter 11 – Radiation Protection Evaluation

Chapter 12 – Accident 

Analysis Evaluation 

• Estimated Dose Rates 
 11-6 

 • The cask system design should satisfy ALARA and other occupational exposure 

requirements of 10 CFR Part 20, and 

 • The sum of the doses from direct radiation and from release of radioactive 

materials to the atmosphere should satisfy the requirements of 10 CFR 72.104(a) 

and 72.106(b). Because of the stringent design requirements for SNF cask 

systems, the release of radionuclides into the atmosphere is expected to be 

insignificant under both normal and accident conditions. Direct radiation is the 

major mode of exposure. 

11.5.2 Occupational Exposures (MEDIUM Priority) 

The reviewer should analyze Chapter 9, “Operating Procedures,” of the SAR and direct 

radiation dose calculations in Chapter 6, “Shielding Evaluation” of the SAR. These data should 

be used in Chapter 11, “Radiation Protection” of the SAR to estimate the doses received by 

occupational personnel, including minors, during cask loading and transfer to the ISFSI. Any 

significant differences from these doses that may occur during cask retrieval and unloading 

should be identified. In addition, the reviewer should verify that the applicant presents similar 

dose estimates for periodic or routine maintenance as well as surveillance activities. These 

estimates may require additional assumptions concerning adjacent casks for a typical storage 

configuration. 

The reviewer should verify that the applicant presents the rationale used to justify the bases for 

various exposure times, personnel locations relative to the casks (including hot spots), number 

of personnel required, and appropriate gamma and neutron dose rates. In addition, the 

reviewer should verify that the calculated doses are consistent with these estimates. The actual 

operations will be performed under an active dose-monitoring program that further ensures 

compliance with the requirements of 10 CFR Part 20. Regulatory Guide (RG) 8.34, “Monitoring 

Criteria and Methods to Calculate Occupational Radiation Doses,” which was developed to 

implement revisions to 10 CFR Part 20, can be used to determine the acceptability of the 

applicant’s occupational exposure evaluation and monitoring recommendations. 

11.5.3 Exposures at or Beyond the Controlled Area Boundary (MEDIUM Priority) 

As required by 10 CFR 72.236(d), the application must demonstrate that the shielding and 

confinement features of the cask are sufficient to meet the requirements for real individuals in 

10 CFR 72.104, and for DBA conditions in 10 CFR 72.106. These demonstrations in the 

application facilitate future site-specific evaluations for each general ISFSI licensee. The real 

individual is an individual at or beyond the controlled area. Dose to any real individual must not 

exceed the limits specified in 10 CFR 72.104 from both the storage facility and other 

surrounding fuel cycle activities. For example, a real individual may be anyone living, working, 

or recreating close to the facility for a significant portion of the year. 

However, for approval of a cask design, the reviewer should ensure that the applicant evaluates 

the shielding and confinement features of a single cask and a theoretical array of casks, 

assuming design-basis source terms and full-time occupancy. Supplemental shielding that may 

be used at an ISFSI to meet the exposure requirements to a real individual should also be 

appropriately evaluated. The reviewer should coordinate the review of supplemental shielding 

with the Chapter 13, “Technical Specifications and Operating Controls and Limits Evaluation,” of 

this SRP review. 
 11-7 

11.5.3.1 Normal Conditions 

The single-cask analysis should identify the minimum distance that is required to meet the 

doses in 10 CFR 72.104. Past applications have shown this distance to be typically within 

200m of the cask. A dose or dose rate versus distance curve for a single cask should be 

included to facilitate site-specific evaluations for general ISFSI licensees. To satisfy section 10 

CFR 72.106(b), dose evaluations should be determined at a minimum of 100m (328 ft) distance 

to the closest boundary of the controlled area. However, the applicant may use a longer 

distance provided that the longer distance is made a condition of use. In addition, the SAR 

should determine the degree to which the normal condition dose rates could change for the 

identified off-normal conditions. 

The reviewer should verify that the applicant includes a dose or dose rate versus distance curve 

in its evaluation of offsite dose for a hypothetical cask array. The theoretical cask array should 

consist of at least 20 storage casks (2x10 array), and the analysis may include the effect of 

shielding among casks in the array. The reviewer should examine predicted dose rates and 

compare them to the dose rates from previously approved casks, and any associated annual 

doses that have been observed for the casks at existing ISFSIs. 

It is important to note that the general ISFSI licensee is permitted to use either distance 

between the ISFSI and the controlled area boundary or engineered features (supplemental 

shielding) such as berms to mitigate doses to real individuals near the site. The SAR needs to 

provide sufficient information to support informed choices on the part of the general licensee. If 

the SAR analyses were performed for the minimum 100-meter distance and did not use any 

additional shielding, and the projected dose at 100 meters exceeded the regulatory limits, the 

reviewer should verify that the application contains a justification for how a general licensee 

could reasonably meet the requirements of Section 72.104. If the dose versus distance curves 

for the single cask and hypothetical array in the SAR were only evaluated at distances greater 

than 100 m, or assumed some engineered feature, then the CoC should contain a condition of 

use to that effect. 

An example of such a condition may be similar to the following: “The use of this system may 

require more than the minimum 100-meter distance between the ISFSI and the controlled area 

boundary, or engineered features (i.e., berms or shield walls), or both to ensure the dose limits 

in 10 CFR 72.104 can be met. In cases where engineered features are used to ensure that the 

requirements of 10 CFR 72.104(a) are met, such features are to be considered important to 

safety [ITS] and must be evaluated to determine the applicable [QA] category.” 

If an engineered feature is used in the SAR evaluations, then that feature is to be considered to 

be part of the system. As such, it should be described in the CoC. 

As required by 72.212(b)(2)(i)(C), a general licensee must perform a written evaluation to 

demonstrate that the requirements of 72.104 are met. An evaluation similar to that for a sitespecific ISFSI should be performed. The licensee may use information provided in the cask 

SAR as well as site-specific information to perform the evaluation. Evaluations performed by 

the general ISFSI licensee are not submitted to NRC for approval; however, they are subject to 

NRC inspection and should be recorded and maintained by the general licensee. 

The general licensee should establish measures in the radiological protection program, 

environmental monitoring program, and/or operating procedures to identify and re-evaluate 
 11-8 

potential increases in exposure to the real individuals. Compliance with the dose limits in 

10 CFR 72.104 will be verified by the environmental monitoring program with direct radiation 

measurements and/or effluent measurements, as appropriate. 

11.5.3.2 Accident Conditions and Natural Phenomenon Events 

The direct dose rate associated with accident conditions at the boundary of the controlled area 

should be reviewed as discussed in Chapter 6, “Shielding Evaluation,” of this SRP. Also, the 

dose rate resulting from accidental release of radionuclides, as presented in Chapter 5, 

“Confinement Evaluation,” of this SRP, should be reviewed. The accident-related radionuclide 

release dose should account for both air and liquid pathways as appropriate. In addition, the 

reviewer should verify that the applicant has evaluated the source terms for both SNF fission 

product and cask surface contamination. The sum of these should satisfy the requirements of 

10 CFR 72.106(b). For purposes of demonstrating compliance with 10 CFR 72.106(b) and 

evaluation against the Environmental Protection Agency Protective Action Guides in the Manual 

of Protective Action Guides and Protective Actions for Nuclear Incidents (EPA 410R-92-001), 

the skin, extremities, and the lens of the eye may be considered separately from other organs. 

As noted in Chapter 6, “Shielding Evaluation,” of this SRP, the time-integrated dose at the 

boundary of the controlled area may be small. Consequently, the reviewer should verify that the 

applicant estimates the doses at 100m (328 ft.) from the storage location to the nearest 

boundary of the controlled area unless the SAR specifies a greater minimum distance that is 

also made a condition of use for the proposed DSS. Alternatively, applicants may depict dose 

estimation using a curve showing dose versus distance from an assumed array of casks. For 

those systems for which these curves indicate the need for greater distance and/or engineered 

features, such as berms, a condition similar to that described in the preceding section of this 

SRP may be needed, with the requirement being 72.106(b) in this instance. 

11.5.4 ALARA (MEDIUM Priority) 

Further information on ALARA can be found in RG 8.8, “Information Relevant to Ensuring that 

Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As is Reasonably 

Achievable,” and RG 8.10, “Operating Philosophy for Maintaining Occupational Radiation 

Exposures As Low As is Reasonably Achievable.” 

11.5.4.1 Design Considerations 

The cask design features should be reviewed to ensure that the features for which credit is 

taken in radiation protection analyses are clearly identified on the drawings. Also, the reviewer 

should ensure the application includes commitments to implement those features that have 

been credited in analyses to show compliance with regulatory requirements or ALARA goals. 

The reviewer should coordinate with the reviewers of SRP Chapters 5, “Confinement 

Evaluation” and 6, “Shielding Evaluation.” 

11.5.4.2 Procedures and Engineering Controls 

The reviewer should determine that the descriptions of proposed DSS operations adequately 

demonstrate that ALARA principles have been incorporated into operational procedures and 

engineering controls. The reviewer should ensure that plans and procedures have been 

developed in accordance with applicable requirements and guidance. 
 11-9 

11.6 Evaluation Findings 

Evaluation findings are prepared by the reviewer upon determination that the regulatory 

requirements related to radiation protection as identified in Section 11.3 of this chapter have 

been satisfied. Some of these determinations can be made only after evaluating the results of 

reviews performed under other chapters of this SRP. If the documentation submitted with the 

application fully supports positive findings for each of the regulatory requirements, the 

statements of findings should be similar to the following: 

 F11.1 The [cask designation] provides radiation shielding and confinement features that 

are sufficient to meet the requirements of 10 CFR 72.104 and 72.106. 

 F11.2 The design and operating procedures of the [cask designation] provide 

acceptable means for controlling and limiting occupational radiation exposures 

within the limits given in 10 CFR 20 and for meeting the objective of maintaining 

exposures ALARA. 

A summary statement similar to the following should be made: 

“The staff concludes that the design of the radiation protection system of the [cask 

designation] is in compliance with 10 CFR Part 72 and that the applicable design and 

acceptance criteria have been satisfied. The evaluation of the radiation protection 

system design provides reasonable assurance that the [cask designation] will allow safe 

storage of SNF. This finding is reached on the basis of a review that considered the 

regulation itself, appropriate regulatory guides, applicable codes and standards, 

accepted health physics practices, and the statements and representations in the 

application.” 
 12-1 

12 ACCIDENT ANALYSES EVALUATION 

12.1 Review Objective 

In this portion of the dry storage system (DSS) review, the U.S. Nuclear Regulatory Commission 

(NRC) evaluates the applicant’s identification and analysis of hazards as well as the summary 

analysis of system responses to both off-normal and accident or design-basis events. 

Normal conditions are the intended operations, planned events, and environmental conditions, 

that are known or reasonably expected to occur with high frequency during storage operations. 

Off-normal events are those man-made events or natural phenomena expected to occur with 

moderate frequency or once per calendar year. ANSI/ANS 57.9 refers to these events as 

Design Event II. 

Design-basis accident events are considered to occur infrequently, if ever, during the lifetime of 

the facility. ANSI/ANS 57.9-92 subdivides this class of accidents into two categories – Design 

Events III and IV. Design Event III is a set of infrequent events that could be expected to occur 

during the lifetime of a DSS, and Design Event IV is a set of events that establishes a 

conservative design basis for structures, systems, and components (SSC) important to safety. 

The effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, 

and seiches, with severity frequencies consistent with Design Event III and IV, are considered to 

be design-basis accident events, in addition to design-basis man-made events. 

This review ensures that the applicant has conducted thorough accident analyses as reflected 

by the following factors: 

 • Identified all credible accidents. 

 • Provided complete information in the safety analysis report (SAR). 

 • Analyzed the safety performance of the cask system in each review area. 

 • Fulfilled all applicable regulatory requirements. 

12.2 Areas of Review 

This portion of the DSS review evaluates the applicant’s identification and analysis of hazards 

with particular emphasis on the safety performance of the cask system under off-normal events 

and conditions, and accident or design-basis events. Consequently, this chapter of the DSS 

Standard Review Plan (SRP) provides guidance for use in reviewing the applicant’s 

identification and analysis of hazards as well as the summary analysis of system responses. A 

comprehensive accident analysis evaluation may encompass the following areas of review: 

Cause of the Event

Detection of the Event

Summary of Event Consequences and Regulatory Compliance

Corrective Course of Action

12.3 Regulatory Requirements 

This section presents a summary matrix of the portions of U.S. Code of Federal Regulations 

(CFR), Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel 
 12-2 

and High-Level Radioactive Waste,” Title 10, “Energy” (10 CFR Part 72) that are relevant to the 

review areas addressed by this chapter. The NRC staff reviewer should read the exact 

referenced regulatory language. Table 12-1 matches the relevant regulatory requirements 

associated with this chapter to the areas of review identified in the previous section. 

Table 12-1 Relationship of Regulations and Areas of Review 

Areas of 

Review

10 CFR Part 72 Regulations

72.104 (a) 72.106 (b) 72.122(b)(1),(3), (d), (g), 

(h)(4), (i), (l) 72.124(a) 72.236(c), (d), (l)

Cause of the 

Event ● 

Detection of the 

Event ● ● 

Summary of 

Event 

Consequences 

and Regulatory 

Compliance 

● ● ● ● ●

Corrective 

Course of Action ● 

12.4 Acceptance Criteria 

Accidents and natural phenomena events may share common regulatory and design limits. 

Consequently, the following sections sometimes refer to these scenarios collectively as accident 

conditions. 

By contrast, off-normal conditions (anticipated occurrences) are distinguished, in part, from 

accidents or natural phenomena by the appropriate regulatory guidance and design criteria. For 

example, the radiation dose from an off-normal event must not exceed the limits specified in 

10 CFR Part 20, “Standards for Protection Against Radiation,” and 10 CFR 72.104(a), whereas 

the radiation dose from an accident or natural phenomenon must not exceed the specifications 

of 10 CFR 72.106(b). Accident conditions may also have different allowable structural criteria. 

In general, this portion of the DSS review seeks to ensure that the DSS design and the 

applicant’s hazard identification and analyses of related system responses fulfill the following 

acceptance criteria: 

12.4.1 Dose Limits for Off-Normal Events 

During normal operations and off-normal conditions, the requirements specified in 10 CFR 

Part 20 must be met. In addition, the annual dose equivalent to any individual located beyond 

the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv 

(75 mrem) to the thyroid, and 0.25 mSv (25 mrem) to any other organ as a result of exposure to 

the following sources (10 CFR 72.104): 
 12-3 

 • Planned discharges to the general environment of radioactive materials (with the 

exception of radon and its decay products). 

 • Direct radiation from operations of the ISFSI. 

 • Any other cumulative radiation from uranium fuel cycle operations (i.e., nuclear 

power plant) in the affected area. 

12.4.2 Dose Limit for Design-Basis Accidents 

The dose from any credible design basis accident to any individual located on or beyond the 

nearest boundary of the controlled area may not exceed the limits specified in 10 CFR 72.106. 

Specifically, these are: the more limiting of a total effective dose equivalent of 0.05 Sv (5 rem), 

or the sum of the deep dose equivalent to and the committed dose equivalent to any individual 

organ or tissue (other than the lens of the eye) of 0.5 Sv (50 rem); a lens dose equivalent of 

0.15 SV (15 rem); and a shallow dose equivalent to skin or any extremity of 0.5 Sv (50 rem). 

12.4.3 Criticality 

The spent nuclear fuel (SNF) must be maintained in a subcritical condition under credible 

conditions (i.e., keff, including all biases and uncertainties, equal to or less than 0.95). At least 

two unlikely, independent, and concurrent or sequential changes in the conditions essential to 

nuclear criticality safety should occur before a nuclear criticality accident is deemed to be 

possible (double contingency). 

12.4.4 Confinement 

The cask and its systems important to safety must be evaluated using appropriate tests or by 

other means acceptable to the NRC to demonstrate that they will reasonably maintain 

confinement of radioactive material under credible accident conditions. 

12.4.5 Recovery and Retrievability 

Recovery is the capability to return the stored radioactive material to a safe condition after an 

accident event without endangering public health and safety. This generally means ensuring 

that any potential release of radioactive materials to the environment or radiation exposures is 

not in excess of the limits in 10 CFR Part 20 during post accident recovery operations. 

Retrievability is specified in 10 CFR 72.122(l) and requires that storage systems must be 

designed to allow ready retrieval of spent fuel, high-level radioactive waste, and reactor-related 

GTCC waste for further processing or disposal. Ready retrieval is the ability to move a canister 

containing spent fuel to either a transportation package or to a location where the spent fuel can 

be removed. Ready retrieval also means maintaining the ability to handle individual or canned 

spent fuel assemblies by the use of normal means. Retrievability applies to normal conditions 

and off-normal events, and not to design-basis accident events. 

12.4.6 Instrumentation 

The SAR must identify all instruments and control systems that must remain operational under 

accident conditions. 
 12-4 

12.5 Review Procedures 

Introduction

Figure 12-1 presents an overview of the evaluation process and can be used as a guide to 

assist in coordinating between the review disciplines. 

Figure 12-1 Overview of Accident Analysis Evaluation

Chapter 2 – Principal 

Design Criteria Evaluation

Chapter 9 – Operating 

Procedures Evaluation

• Design Criteria for Safety 

Protection Systems

• Operating Procedures

• Acceptance Test and 

Maintenance

• Material Compatibility

• Corrective Actions

Chapter 4 – Thermal 

Evaluation

• Accident Conditions

Cause of Event

Detection of Event

Summary of Event Consequences and Regulatory Compliance

Corrective Course of Action

Chapter 12 – Accident Analysis Evaluation

Chapter 3 – Structural 

Evaluation

• Off-Normal Conditions

• Accident-Level Events and 

Conditions (Design -Basis 

Events)

• Cause of Event

• Detection of Event

• Retrievability

Chapter 4 – Thermal 

Evaluation

• Model Specifications

• Temperatures

• Pressures

Chapter 11 – Radiation 

Protection Evaluation

• Estimated Dose Rates

• Spent Fuel Cladding

• Material Properties

Chapter 5 – Confinement 

Evaluation

• Design Criteria

• Design Features

• Containment Monitoring 

Capability

• Nuclides with Potential for 

Release

• Off-Normal Conditions

• Leakage of One Seal

• Accident Conditions and 

Natural Phenomenon 

Events

Chapter 6 – Shielding 

Evaluation

• Shielding Model 

Specification

• Displacement

• Combustion

• Decomposition

• Dehydration

• Melting

Chapter 7 – Criticality 

Evaluation

• Criticality Design Criteria 

and Features

• Fuel Specification

• Configuration

Chapter 1 – General 

Information Evaluation

• DSS Design Information

• DSS Description

• Accident Identification

Chapter 8 – Materials 

Evaluation

Chapter 2 – Principal 

Design Criteria Evaluation

Chapter 9 – Operating 

Procedures Evaluation

• Design Criteria for Safety 

Protection Systems

• Operating Procedures

• Acceptance Test and 

Maintenance

• Material Compatibility

• Corrective Actions

Chapter 4 – Thermal 

Evaluation

• Accident Conditions

Cause of Event

Detection of Event

Summary of Event Consequences and Regulatory Compliance

Corrective Course of Action

Chapter 12 – Accident Analysis Evaluation

Chapter 3 – Structural 

Evaluation

• Off-Normal Conditions

• Accident-Level Events and 

Conditions (Design -Basis 

Events)

• Cause of Event

• Detection of Event

• Retrievability

Chapter 4 – Thermal 

Evaluation

• Model Specifications

• Temperatures

• Pressures

Chapter 11 – Radiation 

Protection Evaluation

• Estimated Dose Rates

• Spent Fuel Cladding

• Material Properties

Chapter 5 – Confinement 

Evaluation

• Design Criteria

• Design Features

• Containment Monitoring 

Capability

• Nuclides with Potential for 

Release

• Off-Normal Conditions

• Leakage of One Seal

• Accident Conditions and 

Natural Phenomenon 

Events

Chapter 6 – Shielding 

Evaluation

• Shielding Model 

Specification

• Displacement

• Combustion

• Decomposition

• Dehydration

• Melting

Chapter 7 – Criticality 

Evaluation

• Criticality Design Criteria 

and Features

• Fuel Specification

• Configuration

Chapter 1 – General 

Information Evaluation

• DSS Design Information

• DSS Description

• Accident Identification

Chapter 8 – Materials 

Evaluation
 12-5 

The review procedures presented here describe general procedures for reviewing a DSS 

submittal. The review procedures in Chapter 15 of NUREG-1567, “Standard Review Plan for 

Spent Fuel Dry Storage Facilities,” provide more detailed procedures and, where applicable, 

may be used as a guide to supplement the review procedures presented herein. 

The off-normal conditions, accidents, and natural phenomena events identified in SAR 

Chapter 2, “Principal Design Criteria” should be reviewed by all disciplines, especially those 

accidents with potential consequences resulting in the failure of the confinement boundary. Offnormal conditions should be evaluated against the requirements of 10 CFR 72.104. Accidents 

and natural phenomena events should be evaluated against the requirements of 10 CFR 72.106 

and 72.122(b). Recovery methods or the need for overpacks or dry transfer systems to 

maintain safe storage conditions would then not be considered and evaluated as part of the 

NRC approval process. For each type of event, this discussion should include the applicant’s 

evaluation of the following areas, as applicable. 

12.5.1 Cause of the Event (MEDIUM Priority) 

The cause of the accident should be described. The description should include the chain of 

events that leads to the credible accident condition and any bounding conditions. 

12.5.2 Detection of the Event (MEDIUM Priority) 

The licensee may detect an event through surveillance programs or monitoring instrumentation 

and alarms. Surveillance programs and monitoring instrumentation and alarms should have 

reasonable flexibility to allow for the identification of an accident condition or noncompliance 

situation that has not been previously considered in the SAR. The method of detection will be 

intuitively obvious for some events, whereas other events (e.g., fuel rod rupture) may remain 

undetected for a significant period of time. 

DSS monitoring equipment (such as a pressure monitoring system) are classified as not 

important to safety, but are classified as Category B under the guideline of NUREG/CR-6407, 

“Classification of Transportation Packaging and Dry Spent Fuel Storage Components According 

to Importance to Safety.” Reviewers should refer to Chapter 5, “Confinement Evaluation,” of 

this SRP. 

12.5.3 Summary of Event Consequences and Regulatory Compliance 

(MEDIUM PRIORITY) 

The applicant should address event consequences in each functional area corresponding to 

earlier chapters of the SAR (i.e., structural, thermal, shielding, criticality, confinement, materials, 

and radiation protection). This discussion should refer back to each SAR chapter in which the 

individual consequences are evaluated in detail. The applicant should provide a summary of 

the accident dose calculations and show that the consequences comply with the applicable 

regulatory criteria. For off-normal conditions, the applicant should demonstrate compliance with 

Part 20 as well as Part 72. 

12.5.4 Corrective Course of Action (MEDIUM Priority) 

The applicant should identify what action(s), if any, would be necessary to recover from the 

event. If various courses of action are possible, the applicant should present a discussion 
 12-6 

concerning the selection of the most appropriate action. Because the fuel must be readily 

retrievable, returning the cask to the fuel handling building and reloading the SNF into a new 

cask is a viable option. If corrective courses of action are to be included in operating 

procedures or administrative programs, then the applicable sections of SAR Chapter 9, 

“Operating Procedures,” should be referenced. 

12.6 Evaluation Findings 

Review the 10 CFR Part 72 acceptance criteria and provide a summary statement for each. 

These statements should be similar to the following model: 

 F12.1 Structures, systems, and components of the [cask designation] are adequate to 

prevent accidents and to mitigate the consequences of accidents and natural 

phenomena events that do occur. 

 F12.2 The spacing of casks, discussed in Chapter of the safety evaluation 

report (SER) and included as an operating limit in Chapter 13, “Technical 

Specifications and Operation Controls and Limits Evaluation” of the SAR will 

ensure accessibility of the equipment and services required for emergency 

response. 

 F12.3 Table of the SER lists the Technical Specifications for the [cask system 

designation]. These Technical Specifications are further discussed in 

Chapter ___ of the SER. 

 F12.4 The applicant has evaluated the [cask designation] to demonstrate that it will 

reasonably maintain confinement of radioactive material under credible accident 

conditions. 

 F12.5 An accident or natural phenomena event will not preclude the ready retrieval of 

SNF for further processing or disposal. 

 F12.6 The SNF will be maintained in a subcritical condition under accident conditions. 

 F12.7 Neither off-normal nor accident conditions will result in a dose to an individual 

outside the controlled area that exceeds the limits of 10 CFR 72.104(a) or 

72.106(b), respectively. 

 F12.8 No instruments or control systems are required to remain operational under 

accident conditions [as applicable]. 

The reviewer should provide a summary statement similar to the following: 

“The staff concludes that the accident design criteria for the [DSS designation] are in 

compliance with 10 CFR Part 72, and the accident design and acceptance criteria have 

been satisfied. The applicant’s accident evaluation of the cask adequately demonstrates 

that it will provide for safe storage of SNF during credible accident situations. This 

finding is reached on the basis of a review that considered independent confirmatory 

calculations, the regulation itself, appropriate regulatory guides, applicable codes and 

standards, and accepted engineering practices.” 
 13-1 

13 TECHNICAL SPECIFICATIONS AND OPERATING CONTROLS AND LIMITS 

EVALUATION 

13.1 Review Objective 

The technical specifications and operating controls and limits review ensures that the operating 

controls and limits or the technical specifications, including their bases and justification, meet 

the requirements of the U.S. Code of Federal Regulations (CFR), Part 72, “Licensing 

Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive 

Waste and Reactor-Related Greater Than Class C Waste,” Title 10, “Energy” (10 CFR Part 72). 

This evaluation is based on information that the applicant presents in Safety Analysis Report 

(SAR) Chapter 13, “Technical Specifications and Operation Controls and Limits Evaluation” as 

well as accepted practices and the applicant’s commitments discussed in other chapters of the 

SAR or in correspondence subsequent to submission of the application. The NRC staff should 

also describe in the Safety Evaluation Report (SER) any additional operating controls and limits 

that the staff deems necessary and has added them, as appropriate, to the cask system’s 

Technical Specifications. 

For simplicity in defining the acceptance criteria and review procedures, the term “technical 

specifications” may be considered synonymous with “operating controls and limits.” The 

technical specifications define the conditions that are deemed necessary for safe dry storage 

system (DSS) use. Specifically, they define operating limits and controls, monitoring 

instruments and control settings, surveillance requirements, design features, and administrative 

controls that ensure safe operation of the DSS. As such, these technical specifications are 

included in a DSS Certificate of Compliance (CoC). Each specification should be clearly 

documented and justified in the technical evaluation sections of the SAR and the associated 

SER as necessary for safe DSS operation. 

If a reviewer determines that a design feature, content specification, analytical assumption, 

operating assumption, limiting condition of operation, element of reactor programmatic controls, 

or other SAR item is important and should not be changed without NRC staff approval, then it 

should be further evaluated and considered as a potential CoC condition or technical 

specification. The reviewer should consider, in part, risk-insights, safety margins, operational 

experience, defense-in-depth considerations, design novelty, and other issues that are unique 

to each proposed design. The reviewer should also implement the guidance in this chapter for 

establishing such conditions and technical specifications in the CoC. 

13.2 Areas of Review 

This chapter of the DSS Standard Review Plan (SRP) provides guidance for use in evaluating 

the technical specifications that the applicant deems necessary for safe use of the proposed 

DSS system. As defined in Section 13.5, “Review Procedures,” a comprehensive review of the 

proposed technical specifications would assess the applicant’s compliance with the regulations 

to provide a level of control commensurate with that specified by 10 CFR 72.234 and 72.236. 

These requirements represent the following areas of review: 

Functional/Operating Limits, Monitoring Instruments, and Limiting Control 

Settings

Limiting Conditions
 13-2 

Surveillance Requirements

Design Features

Administrative Controls

13.3 Regulatory Requirements 

This section presents a summary matrix of the portions of 10 CFR Part 72 that are relevant to 

the review areas addressed by this chapter. The U.S. Nuclear Regulatory Commission (NRC) 

staff reviewer should read the exact referenced regulatory language. Table 13-1 matches the 

relevant regulatory requirements associated with this chapter to the areas of review identified in 

the previous section. 

Table 13-1 Relationship of Regulations and Areas of Review 

Areas of Review

10 CFR Part 72 Requirements

72.234 

(a) 

72.236 

(a) (b) (c) (d) 

(e), 

(f), 

(h) 

(g) (i) (j) (l) 

Functional/Operating Limits, Monitoring 

Instruments, and Limiting Control 

Settings 

● ● ● ● ●

Limiting Conditions ● ● ● ● ●

Surveillance Requirements ● ● ● ● 

Design Features ● ● ● ● ● ● ●

Administrative Controls ● ● ● ● ●

13.4 Acceptance Criteria 

The reviewer should verify that the applicant identifies proposed technical specifications 

necessary to maintain subcriticality, confinement, shielding, heat removal, and structural 

integrity under normal, off-normal, and accident-level conditions. In addition, the reviewer 

should ensure that the applicant identifies the basis for each of the proposed technical 

specifications by reference to the analysis in the SAR. The NRC staff can use NUREG-1745, 

“Standard Format and Content for Technical Specifications for 10 CFR Part 72 Cask Certificates 

of Compliance,” as an appropriate template in the review of the technical specifications. 

However, the staff may impose alternative technical specifications to NUREG-1745 guidance, 

based on operational experience, and the Office of General Counsel legal interpretations that 

have been made since issuance of NUREG-1745. 
 13-3 

13.4.1 Functional/Operating Limits, Monitoring Instruments, and Limiting Control 

Settings 

Acceptance criteria for functional and operating limits, monitoring instruments, and limiting 

control settings include limits placed on fuel, waste handling, and storage conditions to protect 

the integrity of the fuel and container, to protect the employees against occupational exposures, 

and to guard against the uncontrolled release of radioactive materials. 

13.4.2 Limiting Conditions 

Acceptance criteria for functional and operating limits, monitoring instruments, and limiting 

control settings include limits placed on fuel, waste handling, and storage conditions to protect 

the integrity of the fuel and container, to protect the employees against occupational exposures, 

and to guard against the uncontrolled release of radioactive materials. Acceptance criteria for 

limiting conditions are the lowest levels required for safe operation. 

13.4.3 Surveillance Requirements 

Acceptance criteria for establishing surveillance requirements include the frequency and scope 

of surveillance requirements to verify performance and availability of structures, systems, and 

components (SSCs) important to safety, and the verification of the bases for the proposed 

limiting conditions. 

13.4.4 Design Features 

Acceptance criteria for design features include commitments to specified codes. The condition 

or technical specification should also describe a process to address deviations from the 

applicable codes that may be necessary. In such cases, the licensee should request an 

alternative to the requirements of the applicable code from the NRC. If the staff finds that the 

deviation does not adversely impact safety, it may authorize the requested alternative in writing. 

Currently, there is an existing code for the design and construction of metallic nuclear fuel 

storage casks and the document is identified as Subsection WC of Division 3 of Section III of 

the ASME Boiler and Pressure Vessel Code. This was first issued as the 2005 addenda to the 

2004 Code. The current Code edition is 2007. As of February 2008, NRC staff had not taken a 

position regarding the acceptability of this document. In the past, Division 1 of the ASME B&PV 

Code had been used by NRC staff allowing alternatives to some provisions of that document 

which were judged to not be applicable to spent nuclear fuel storage casks. Early SNF dry 

storage licenses and certificates of compliance were issued without documenting which specific 

alternatives to ASME B&PV Code, Section III, were approved. Poor quality assurance practices 

during design and fabrication sometimes led to significant deviations from the Code without 

appropriate certificate holder design review or NRC review and approval. Therefore, the 

applicant should document commitments to ASME B&PV Code, Section III, with proposed 

alternatives in the application. 

Likewise the NRC should document these commitments in the 10 CFR Part 72 licenses, 

certificates of compliance, or technical specifications and its approval of the proposed 

alternatives in the SER. Also, the NRC should include a statement (in the CoC or technical 

specifications) that refers the reader to the SAR and applicable SERs for any alternatives to the 

codes. In addition, to ensure that similar problems do not exist in other areas, all other codes 

and standards applied to components important to safety should be identified in the SAR and 
 13-4 

should be included in the CoC or technical specifications. Figure 13-1 presents an example of a 

provision for allowing alternatives to applicable codes. 

Figure 13-1 Provision Example 

In addition, acceptance criteria for design features include specifications important to criticality 

safety. Where criticality analyses rely upon the condition that the assemblies’ active fuel length 

remains within the cask region containing the solid neutron absorbers, the applicant should 

commit to ensuring the cask features fulfill this analysis assumption. One common method is 

the installation of fuel spacers, upper and/or lower spacers as needed, to maintain the 

assemblies’ position under all credible conditions. The minimum Boron-10 content of the solid 

neutron absorbers is another important design feature specification together with the 

qualification and acceptance testing method for ensuring the neutron absorbers meet the 

required minimum Boron-10 content throughout the absorber material. The proximity of fuel 

assemblies to each other also affects the cask’s reactivity, generally with reactivity increasing as 

the assemblies are brought closer together; therefore, a minimum dimension(s) between 

adjacent assembly locations is specified. This dimension may be a minimum flux trap width or a 

minimum fuel cell pitch. These design parameters and commitments should also be included in 

the license, certificate of compliance, or technical specifications. 

#.#.# Codes and Standards

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel 

(B&PV) Code, Section III, 1992 Edition with Addenda through 1994 is the governing 

Code for the storage system. 

#.#.#.# Design Alternatives to Codes, Standards, and Criteria

 Table #-# lists all approved alternatives for the design of the DSS. 

#.#.#.# Construction/Fabrication Alternatives to Codes, Standards, and Criteria

Proposed alternatives to ASME B&PV Code Section III, 1992 Edition with Addenda 

through 1994, including alternatives referenced in Section 4.3.1, may be used when 

authorized by the Director of the Office of Nuclear Material Safety and Safeguards or 

designee. 

The proposal to the NRC must demonstrate that the alternatives would provide an 

acceptable level of quality and safety, or that compliance with the specified 

requirements of ASME B&PV Code, Section III, 1992 Edition with Addenda through 

1994 would result in hardship or unusual difficulty without a compensating increase 

in the level of quality and safety. 
 13-5 

13.4.5 Administrative Control 

Acceptance criteria for administrative controls include organizational and management 

procedures, recordkeeping, review and audit systems, and reporting necessary to ensure that 

the DSS is managed in a safe and reliable manner. Administrative action that must be taken in 

the event of noncompliance with a limit or condition should be specified. 

13.5 Review Procedures (HIGH Priority) 

Figure 13-2 presents an overview of the evaluation process and can be used as a guide to 

assist in coordinating between review disciplines. 

Reviewers should evaluate each chapter of the SAR with the goal of establishing the technical 

specifications. The variability of designs and operations makes it impossible to define each 

instance for which a technical specification is necessary. For this reason, it is important that the 

NRC staff conduct a coordinated, detailed, and thorough evaluation of each technical section of 

the SAR. Reviewers should note all instances in which the SAR either makes an assumption or 

imposes a condition that should be identified as a technical specification. Reviewers should 

also note any instances in which the SAR requests alternatives or exemptions from regulatory 

requirements, or other conditions that the reviewer identifies as an operational limit or condition. 

Such limits and exemptions should be clearly identified and documented in SAR Chapter 13. 

“Technical Specifications and Operation Controls and Limits Evaluation”. 

The various technical disciplines should review the results of their specific evaluations and 

compare their list of technical specifications to those identified by the applicant. The NRC staff 

should ensure that the conditions for use, as evaluated and approved by the technical 

reviewers, complement one another and are not contradictory. In addition, the staff will 

coordinate the resolution of any disputed condition, limit, or specification. The staff is 

responsible for identifying any unique specifications (e.g., administrative) that may not be 

covered in the technical sections, although input may be solicited from the technical reviewers 

regarding any topic. 

All reviewers should be familiar with the technical specifications of similar cask designs 

previously approved by the NRC staff. For example, the staff has previously approved cask 

designs and issued technical specifications regarding a variety of items including, but not limited 

to, the following examples: 

 • General requirements and conditions regarding site-specific parameters, 

operating procedures, quality assurance, heavy loads, training, etc. 

 • A preoperational training exercise and demonstration of most cask operations 

including loading, sealing, and drying (using mockups as appropriate); placement 

in storage; and return of fuel to the SNF pool. 
 13-6 

Figure 13-2 Overview of Technical Specifications and Operating Controls and Limits 

Evaluation

• Operation Monitoring 

and Safety Limits

• Loading, Unloading, and 

Handling

Chapter 2 – Principal 

Design Criteria Evaluation

Chapter 9 – Operating 

Procedures Evaluation

• Components/Features

• Design Basis Source Term

• Monitoring Instrumentation

• Codes and Standards

• Operating Control 

Limits

• Administrative 

Programs

• Surveillance 

Requirements

Surveillance Requirements

Design Features

Administrative Controls

Functional Operating Limits, Monitoring

Instruments, and Limiting Control Settings

Limiting Conditions

Chapter 13 – Technical Specifications and Operating Controls Evaluation

Chapter 3 – Structural 

Evaluation

• Accident Analyses

• Design and Construction

• Cask Handling, 

Operations, and 

Inspections

• Component and Content 

Specifications

Chapter 4 – Thermal 

Evaluation

• Cladding Heating Limits

• Minimum Heat Load from 

Fuel and Cooling Time

Chapter 5 –

Confinement Evaluation

• Leak Tests

• Interseal Pressure

Chapter 6 – Shielding 

Evaluation

• Supplemental Shielding

Chapter 7 – Criticality 

Evaluation

• Burnable Neutron 

Minimum Absorber 

Loading

• Spent Fuel 

Specifications

• Minimum Flux Trap/Fuel 

Cell Pitch

• Fuel Spacers

• Neutron Absorber Tests

Chapter 1 – General 

Information Evaluation

• 10 CFR Part 72 

Requirements

• Components/Systems 

Proposed Operations

• Spent Fuel Specifications 

and Configuration

Chapter 8 – Materials 

Evaluation

Chapter 6 – Shielding 

Evaluation

• Dose Rate Limits

Chapter 11 – Radiation 

Protection Evaluation

• Supplemental Shielding

• Surface Contamination 

and Dose Rate Limits

Chapter 12 – Accident 

Analyses Evaluation

• Listing of Technical 

Specifications

• Operation Monitoring 

and Safety Limits

• Loading, Unloading, and 

Handling

Chapter 2 – Principal 

Design Criteria Evaluation

Chapter 9 – Operating 

Procedures Evaluation

• Components/Features

• Design Basis Source Term

• Monitoring Instrumentation

• Codes and Standards

• Operating Control 

Limits

• Administrative 

Programs

• Surveillance 

Requirements

Surveillance Requirements

Design Features

Administrative Controls

Functional Operating Limits, Monitoring

Instruments, and Limiting Control Settings

Limiting Conditions

Chapter 13 – Technical Specifications and Operating Controls Evaluation

Chapter 3 – Structural 

Evaluation

• Accident Analyses

• Design and Construction

• Cask Handling, 

Operations, and 

Inspections

• Component and Content 

Specifications

Chapter 4 – Thermal 

Evaluation

• Cladding Heating Limits

• Minimum Heat Load from 

Fuel and Cooling Time

Chapter 5 –

Confinement Evaluation

• Leak Tests

• Interseal Pressure

Chapter 6 – Shielding 

Evaluation

• Supplemental Shielding

Chapter 7 – Criticality 

Evaluation

• Burnable Neutron 

Minimum Absorber 

Loading

• Spent Fuel 

Specifications

• Minimum Flux Trap/Fuel 

Cell Pitch

• Fuel Spacers

• Neutron Absorber Tests

Chapter 1 – General 

Information Evaluation

• 10 CFR Part 72 

Requirements

• Components/Systems 

Proposed Operations

• Spent Fuel Specifications 

and Configuration

Chapter 8 – Materials 

Evaluation

Chapter 6 – Shielding 

Evaluation

• Dose Rate Limits

Chapter 11 – Radiation 

Protection Evaluation

• Supplemental Shielding

• Surface Contamination 

and Dose Rate Limits

Chapter 12 – Accident 

Analyses Evaluation

• Listing of Technical 

Specifications
 13-7 

 • Specifications for the SNF to be stored in the cask, including, but not limited to, 

the type of SNF (i.e., boiling water reactor [BWR], pressurized water reactor 

[PWR], or both), the minimum and maximum allowable enrichments of the fuel 

before irradiation, burnup (i.e., megawatt-days/MTU), the minimum acceptable 

cooling time of the SNF before storage in the cask, the maximum heat designed 

to be dissipated, the maximum SNF loading limit, condition of the SNF (i.e., intact 

assembly or consolidated fuel rods, allowable cladding condition), associated 

non-fuel hardware, and physical parameters (e.g., length, width, depth, weight, 

etc.). The reviewer should be aware that additional SNF specifications regarding 

operational history parameters (e.g., average moderator temperature, average 

in-core soluble boron concentrations, and operations under control rod banks or 

with control rod insertion) will need to be included in the technical specifications 

for cask systems relying on burnup credit 

 • Criticality controls such as cask water boron concentrations, minimum flux 

trap/fuel cell pitch, use of fuel spacers, minimum neutron absorber loading, and 

neutron absorber tests. 

 • The inerting atmosphere requirements during vacuum drying and helium backfill 

parameters. 

 • Cask handling restrictions such as lift height limits and ambient temperature 

(high/low) conditions. 

 • Confinement barrier requirements such as helium leak rate limits. 

 • Thermal performance parameters such as maximum temperatures or deltatemperatures. 

 • Radiological controls such as radiation dose rates and contamination limits. 

 • Cask array and/or spacing limits for thermal performance and radiological 

considerations. 

• Definition of damaged fuel 

• Code of record and alternatives to specific Code requirements 

• Specification/requirements for alternative materials for ITS components 

• Manufacture and testing of neutron poison material(s) for criticality control 

• Hydrogen monitoring/mitigation during wet loading/unloading 

• Maintaining inert atmosphere during canister draining/flooding to prevent 

Oxidation 

• Use of copper bearing or weathering steel for structural steel components at 

coastal marine ISFSI sites (or other corrosion mitigation measures) 
 13-8 

• Operational controls to maintain cladding temperature limits 

• Low Temperature Ductility of Ferritic Steels 

All disciplines should coordinate their review of the proposed technical specifications to assure 

the operational limitations are measurable and inspectible. Other topics may include: 

 • Frequency and scope proposed for the surveillance requirements. 

 • Administrative controls that include organization and administrative systems and 

procedures, record-keeping, review, and audit systems required to ensure that 

the DSS is managed in a safe and reliable manner. 

 • Administrative action that must be taken in the event of noncompliance with a 

limit or condition. 

The reviewer should verify that the applicant includes a written description in a condition to the 

CoC or technical specification that documents the codes to which the applicant has committed. 

In addition, the condition or technical specification should describe a process to address any 

deviations from the ASME B&PV Code or other codes that may be needed. Likewise, the 

reviewer should verify that these commitments are documented in the 10 CFR Part 72 CoC or 

technical specifications. A list of proposed alternatives to code requirements should also be 

provided in the SAR. This list should be revised as necessary to reflect all NRC-authorized 

alternatives. 

NUREG-1745 provides a recommended format for use by applicants in presenting technical 

specifications. However, this format may not be applicable to all controls. Since the basis for 

the control may be extensively discussed in earlier chapters of the SAR, the applicant may use 

an abbreviated format in SAR Chapter 13. 

Reviewers should ensure that all necessary technical specifications are explicitly delineated in 

SER Chapter 13, “Technical Specifications and Operating Controls and Limits Evaluation,” and 

in the CoC. These delineations typically restate the technical specifications defined in the SAR 

but may be modified or supplemented as the staff deems appropriate. Reviewers should also 

ensure that limits and exemptions requested by the applicant are clearly identified and 

documented in the SER. The staff may prepare a separate table or appendix for SER 

Chapter 13 to explicitly designate the technical specifications that are applicable to the cask. 

Applicable drawings from the SAR should be identified by number and revision. 

13.6 Evaluation Findings 

NRC staff reviewers prepare evaluation findings regarding satisfaction of the regulatory 

requirements related to technical specifications. Evaluation findings developed or included in all 

SER sections relating to technical specifications are also listed in this section. These 

statements should be similar to the following model: 

 F13.1 The staff concludes that the conditions for use for [DSS name] identify necessary 

technical specifications to satisfy 10 CFR Part 72 and that the applicable 

acceptance criteria have been satisfied. The proposed technical specifications 
 13-9 

provide reasonable assurance that the DSS will allow safe storage of SNF. This 

finding is based on the regulation itself, appropriate regulatory guides, applicable 

codes and standards, and accepted practices. The technical specifications 

identified by the applicant include the following: [Reviewer to specify]. 

The reviewer should provide a summary statement similar to the following: 

“The proposed technical specifications provide reasonable assurance that the cask will 

allow safe storage of spent fuel. This finding is reached on the basis of a review that 

considered the regulation itself, appropriate regulatory guides, applicable codes and 

standards, accepted practices and the statements and representations in the 

application.” 
 14-1 

14 QUALITY ASSURANCE EVALUATION 

14.1 Review Objective 

The objective of the review is to determine whether the applicant for a dry storage system (DSS) 

certificate has submitted a quality assurance (QA) program description (QAPD) that 

demonstrates that the applicant's QA program complies with the requirements of 10 CFR Part 

72, Subpart G (Part 72), “Quality Assurance.” 

The basis for that determination is developed from an evaluation of the applicant's high level 

QAPD against the criteria provided in Section 14.4, Review Procedures below, Part 72, and any 

associated information found in the Federal Register since the last rulemaking has been 

completed, as applicable. (Note: The scope of review does not include actual procedures and 

instructions that implement the QA program, but may be described in the QAPD). 

Determination of compliance for the applicant's QA program occurs during NRC inspection 

activities where implementation of the QA plan is evaluated. (Note: The scope of an inspection 

does include actual procedures and instructions that implement the QA program). 

14.2 Areas of Review 

This SRP provides guidance for use by a reviewer to perform an evaluation of a QAPD in terms 

of the 18 criteria defined in 10 CFR Part 72, Subpart G and Section 14.4, “Review Procedures” 

below, and the Federal Register, as applicable. 

14.3 Regulatory Requirements 

This section identifies the reviewer's need to review the exact regulatory language found in 

Part 72 relevant to quality assurance as applied to a DSS. Refer to Subpart G -Quality 

Assurance of 10 CFR Part 72. 

14.4 Acceptance Criteria 

The acceptance criteria below reflect the 18 quality criteria of Part 72, Subpart G. These criteria 

are presented in the form of descriptions of information to be included in the applicant’s QAPD. 

For each criterion shown in Sections 14.5.1 through 14.5.18 of this SRP, examples of measures 

have been provided which may assist the reviewer in determining if the QAPD indicates that it 

meets the applicable criterion. For each of the activities and items identified as important to 

safety, the applicant should identify the applicable QA programmatic elements and include, as 

applicable, provisions for meeting each of the following quality criteria itemized in Section 14.5. 

14.5 Review Procedures (All items in this section are HIGH Priority) 

The purpose of the review is to obtain reasonable assurance that the applicant has developed 

and described a QA program for design, fabrication, construction, testing, operations, 

modification, and decommissioning activities associated with important-to-safety DSS systems, 

structures and components (SSCs). 

It is important that the applicant's QAPD and associated portions of the safety analysis report 

(SAR) provide sufficient detail to enable the reviewer to assess that the applicant has committed 
 14-2 

to comply with the program and the QA program complies with the applicable requirements of 

10 CFR Part 72, Subpart G. If the reviewer determines that sufficient detail does not exist in the 

QAPD, the reviewer should refer to Section 14.6, Evaluation Findings for further direction. If the 

QAPD indicates commitment to follow certain standards, codes, etc., then the reviewer should 

consider the commitments as an integral part of the QA program. 

The reviewer should recognize that application for QA program approval may either be separate 

from the SAR or may exist as a section in the applicant's SAR. Since it is possible that some 

aspects of the QA program are described in various portions of the application (the SAR or a 

submittal separate from the SAR) the reviewer should consider these aspects when evaluating 

the program against the acceptance criteria of Section 14.4. Therefore, if possible, the QAPD 

evaluation should be coordinated with other aspects of the DSS review. Such coordination will 

allow reviewers to derive a more accurate and complete assessment of the applicant's level of 

commitment to the overall QA program, the selection of quality criteria and quality levels, and 

the proposed implementation methods. 

The applicant's QA program may be structured to apply QA measures and controls to all 

activities and items in proportion to their importance to safety, commonly referred to as a graded 

approach. A graded approach for the application of QA should be described in the QAPD by 

adequately assigning appropriate grading classifications and providing an associated 

justification. However, an applicant may choose to apply the highest level of QA and control to 

all activities and items. The QA program should identify the activities and items that are 

important to safety and the degree of their importance. For application of a graded approach, 

the highly important-to-safety activities and items must have a high level of control, while those 

less important may have a lower level of control. If the QA program is graded, the staff should 

be able to conclude that the structure of the graded program is acceptable and that the highest 

levels of QA are applied to those SSCs that are most important to safety. In making 

determinations about the application of QA to those SSCs that are listed in the description as 

important to safety, the reviewer of the QA program description should coordinate with the 

appropriate NRC project manager and associated technical staff to compare those SSCs 

described in other portions of the applicant's submittal. 

If after review, the reviewer finds the QAPD acceptable, the acceptance of the evaluation should 

be documented in the Safety Evaluation Report (SER) for QAPDs submitted as part of a SAR. 

If the applicant's QAPD was submitted prior to the applicant's SAR submittal, the acceptance of 

the evaluation should be documented in a letter to the applicant and if possible included in the 

SER at a later time. In either case, the documentation of the review should include the basis for 

acceptance as noted in the example in Section 14.6 Evaluation Findings. Any 

recommendations for modifications in the application that are required before the application 

can be accepted should be addressed by referring to Section 14.6 for initiation of a request for 

additional information (RAI). 

Figure 14-1 presents an overview of the evaluation process and can be used as a guide to 

assist in coordinating with other review disciplines. 
 14-3 

Figure 14-1 Quality Assurance Evaluation 

14.5.1 Quality Assurance Organization 

The QAPD should describe the structure, interrelationships, and areas of functional 

responsibility and authority for all organizational elements that will perform activities related to 

quality and safety. The following are examples of areas/items that may be addressed to support 

implementation of the quality criteria: 

a. Measures to retain and exercise responsibility for the QA program. The assignment of 

responsibility for the overall QA program in no degree relieves line management of their 

responsibility for the achievement of quality. 

b. Measures to identify and describe the QA functions performed by the applicant's QA 

organization or delegated to other organizations that will provide controls to ensure 

implementation of the applicable elements of the QA criteria. 

c. Measures to provide clear management controls and effective lines of communication 

should exist between the applicant's QA organizations and suppliers to ensure proper 

direction of the QA program and resolution of QA problems. 

d. Measures to identify onsite and offsite organizational elements that will function under 

the purview of the QA program and the lines of responsibility. 

e. Measures to ensure that high-level management is responsible for documenting and 

promulgating the applicant's QA policies, goals, and objectives, and this management 

level should maintain a continuing involvement in QA matters. The application should 

Chapter 9 – Operating 

Procedures Evaluation

• QA Procedures

• Documentation Procedures

• Procurement and Inventory 

Control

• Special Processes

Chapter 10 – Acceptance Tests 

and Maintenance Program 

Evaluation

• Test Controls

• Inspection Criteria

Control of Special Processes

Licensee and Certificate Holder Inspection

Test Control

Control of Measuring and Test Equipment

Handling, Storage, and Shipping Control

Inspection, Test, and Operating Status

Nonconforming Materials, Parts, or Components

Corrective Action

Quality Assurance Records

Audits

Quality Assurance Organization

Quality Assurance Program

Design Control

Procurement Document Control

Instructions, Procedures, and Drawings

Document Control

Control of Purchased Material, Equipment, and Services

Identification and Control of Materials, Parts, and Components

Chapter 14 – Quality Assurance Evaluation

Chapter 1 – General Information 

Evaluation

• Items Important to Safety

Chapter 9 – Operating 

Procedures Evaluation

• QA Procedures

• Documentation Procedures

• Procurement and Inventory 

Control

• Special Processes

Chapter 10 – Acceptance Tests 

and Maintenance Program 

Evaluation

• Test Controls

• Inspection Criteria

Control of Special Processes

Licensee and Certificate Holder Inspection

Test Control

Control of Measuring and Test Equipment

Handling, Storage, and Shipping Control

Inspection, Test, and Operating Status

Nonconforming Materials, Parts, or Components

Corrective Action

Quality Assurance Records

Audits

Quality Assurance Organization

Quality Assurance Program

Design Control

Procurement Document Control

Instructions, Procedures, and Drawings

Document Control

Control of Purchased Material, Equipment, and Services

Identification and Control of Materials, Parts, and Components

Chapter 14 – Quality Assurance Evaluation

Chapter 1 – General Information 

Evaluation

• Items Important to Safety
 14-4 

also describe the lines of communication between intermediate levels of management 

and between this position and the Manager (or Director) of QA. 

f. Measures to designate a position that retains overall authority and responsibility for the 

QA program. 

g. Measures to provide authority and independence of the individual responsible for 

managing the QA program should be such that he or she can direct and control the 

organization's QA program, effectively ensure conformance to quality requirements, and 

remain sufficiently independent of undue influences and responsibilities of schedules 

and costs. In addition, measures to have this individual report to at least the same 

organizational level as the highest line manager directly responsible for performing 

activities affecting quality. 

h. Measures for individuals or groups responsible for defining and controlling the content of 

the QA program and related manuals to have appropriate organizational position and 

authority, as should the management level responsible for final review and approval. 

I. Measures describing the qualification requirements for the principal QA management 

positions so as to demonstrate management and technical competence commensurate 

with the responsibilities of these positions. 

j. Measures to ensure conformance to established requirements be verified by individuals 

or groups who do not have direct responsibility for performing the work being verified. 

The quality control function may be part of the line organization provided that the QA 

organization performs periodic surveillance to confirm sufficient independence from the 

individuals who performed the activities. 

k. Persons and organizations performing QA functions should have direct access to 

management levels that will ensure accomplishment of quality-affecting activities. These 

individuals should have sufficient authority and organizational freedom to perform their 

QA functions effectively and without reservation. In addition, they should be able to 

identify quality problems; initiate, recommend, or provide solutions through designated 

channels; and verify implementation of solutions. 

l. Designated QA individuals or organizations should have the responsibility and authority, 

delineated in writing, to stop unsatisfactory work and control further processing, delivery, 

or installation of nonconforming material. In addition, the application should describe 

how stop-work requests will be initiated and completed. 

m. Measures to determine the extent of QA controls to be determined by the QA staff in 

combination with the line staff and to depend upon the specific activity or item complexity 

and level of importance to safety. 

14.5.2 Quality Assurance Program 

The QAPD should provide acceptable evidence that the applicant's proposed QA program will 

be well-documented, planned, implemented, and maintained to provide the appropriate level of 

control over activities and SSCs consistent with their relative importance to safety. The 

following are examples of areas/items that may be addressed to support implementation of the 

quality criteria: 
 14-5 

a. Measures used to ensure that the QA program meets applicable acceptance criteria. 

b. Measures for management to regularly assess the effectiveness of the QA program. In 

addition, measures for management (above and beyond the QA organization) to 

regularly assess the scope, status, adequacy, and compliance of the QA program to the 

requirements of 10 CFR Part 72. Measures to provide for management's frequent 

contact with program status through reports, meetings, and audits as well as 

performance of a periodic assessment that is planned and documented with corrective 

action identified and tracked. 

c. Measures used to ensure that trained, qualified personnel within the organization will be 

assigned to determine that functions delegated to contractors are properly 

accomplished. 

d. Summarizations of the corporate QA policies, goals, and objectives and establishment of 

a meaningful channel for transmittal of these policies, goals, and objectives down 

through the levels of management. 

e. Measures to designate responsibilities for implementing the major activities addressed in 

the QA manuals. 

f. Measures to control the distribution of the QA manuals and revisions. 

g. Measures for communicating to all responsible organizations and individuals that 

policies, QA manuals, and procedures are mandatory requirements. 

h. Measures to provide a comprehensive listing of QA procedures, plus a matrix of these 

procedures cross-referenced to each of the QA criteria, to demonstrate that the QA 

program will be fully implemented by documented procedures. 

I. Identification of the structures, systems, and components (SSCs) that are important to 

safety and how they will be controlled by the QA program. 

j. Measures for review and documents to show agreement with the QA program provisions 

of its suppliers to ensure implementation of a program meeting the QA criteria. 

k. Measures for the resolution of disputes involving quality arising from a difference of 

opinion between QA/Quality Control (QC) personnel and personnel from other 

departments (engineering, procurement, manufacturing, etc.). 

l. Measures for indoctrination, training, and qualification programs that fulfill the following 

criteria: 

• Personnel responsible for performing activities affecting quality should be 

instructed as to the purpose, scope, and implementation of the quality-related 

manuals, instructions, and procedures. 

• Personnel performing activities affecting quality should be trained and qualified in 

the principles and techniques of the activities being performed. 
 14-6 

• Maintenance of the proficiency of personnel performing quality-affecting activities 

by retraining, reexamining, and re-certifying. 

• Preparation and maintenance of documentation of completed training and 

qualification. 

• Qualification of personnel in accordance with accepted codes and standards. 

14.5.3 Design Control 

The QAPD should describe the approach that the applicant will use to define, control, and verify 

the design and development of the DSS. The following are examples of areas/items that may 

be addressed to support implementation of the quality criteria: 

a. Measures to carry out design activities in a planned, controlled, and orderly manner. 

b. Measures to correctly translate the applicable regulatory requirements and design bases 

into specifications, drawings, written procedures, and instructions. 

c. Measures to describe how the applicant will specify quality standards in the design 

documents and control deviations and changes from these quality standards. 

d. Measures to describe how the applicant will review designs to ensure that design 

characteristics can be controlled, inspected, and tested and that inspection and test 

criteria are identified. 

e. Measures to describe how the applicant will establish both internal and external design 

interface controls. These controls should include review, approval, release, distribution, 

and revision of documents involving design interfaces with participating design 

organizations. 

f. Measures to describe how they will properly select and perform design verification 

processes such as design reviews, alternative calculations, or qualification testing. 

When a test program is to be used to verify the adequacy of a design, the measures 

should be developed to describe how they will use a qualification test of a prototype unit 

under adverse design conditions. 

g. Design verification constitutes confirmation that the design of the SSC is suitable for its 

intended purpose. Measures to ensure design verifications are completed by an 

individual with a level of skill at least equal to that of the original designer, recognizing 

design checking can be performed by a less experienced person. (As an example, 

design checking, which should also be performed, includes confirmation of the numerical 

accuracy of computations and the accuracy of data input to computer codes. 

Confirmation that the correct computer code has been used is part of design 

verification.) Measures to describe how design verification will be performed by persons 

other than those performing design checking. In addition, measures to include how 

individuals or groups responsible for design verification will not include the original 

designer and normally not include the designer's immediate supervisor. 
 14-7 

h. Measures to ensure design and specification changes are subject to the same design 

controls and the same or equivalent approvals that were applicable to the original 

design. 

I. Measures to ensure the documentation of all errors and deficiencies in the design or the 

design process that could adversely affect SSCs important to safety. In addition, the 

applicant should provide measures for adequate corrective action, including root cause 

evaluation of significant errors and deficiencies, to preclude repetition. 

 

j. Before selecting materials, parts, and equipment that are standard, commercial (off-theshelf), or have been previously approved for a different application, measures should be 

provided to review the suitability of any materials, parts, and equipment for the intended 

application. 

k. Measures to provide written procedures to identify and control the authority and 

responsibilities of all individuals or groups responsible for design reviews and other 

design verification activities. 

l. Measures that include the use of valid industry standards and specifications for the 

selection of suitable materials, parts, equipment, and processes for SSCs that are 

important to safety. 

14.5.4 Procurement Document Control 

Documents used to procure SSCs or services should include or reference applicable design 

bases and other requirements necessary to ensure adequate quality. The following are 

examples of areas/items that may be addressed to support implementation of the quality 

criteria: 

a. Measures to establish procedures that clearly delineate the sequence of actions to be 

accomplished in the preparation, review, approval, and control of procurement 

documents. 

b. Measures to ensure that qualified personnel review and concur with the adequacy of 

quality requirements stated in procurement documents. These measures should also 

ensure that the quality requirements are correctly stated, inspectible, and controllable; 

there are adequate acceptance and rejection criteria; and the procurement document 

has been prepared, reviewed, and approved in accordance with QA program 

requirements. 

c. Measures to document the review and approval of procurement documents before they 

are released, and the documentation should be available for verification. 

d. Procurement documents should identify the applicable QA requirements that should be 

compiled and described in the supplier's QA program. In addition, the applicant should 

review and concur with the supplier's QA program. 

e. Measures to ensure procurement documents contain or reference the regulatory 

requirements, design bases, and other technical requirements. 
 14-8 

f. Measures to ensure procurement documents identify the documentation (e.g., drawings, 

specifications, procedures, inspection and fabrication plans, inspection and test records, 

personnel and procedure qualifications, and chemical and physical test results of 

material) to be prepared, maintained, and submitted to the purchaser for review and 

approval. 

g. Measures to ensure procurement documents identify records to be retained, controlled, 

and maintained by the supplier and those records to be delivered to the purchaser 

before use or installation of the hardware. 

h. Measures to ensure procurement documents specify the procuring agency's right of 

access to the supplier's facilities and records for source inspection and audit. 

I. Measures to ensure that changes and revisions to procurement documents are subject 

to the same or equivalent review and approval as the original documents. 

14.5.5 Instructions, Procedures, and Drawings 

The QAPD should define the applicant's proposed procedures for ensuring that activities 

affecting quality will be prescribed by, and performed in accordance with, documented 

instructions, procedures, or drawings of a type appropriate for the circumstances. The following 

are examples of areas/items that may be addressed to support implementation of the quality 

criteria: 

a. Measures to ensure activities affecting quality are prescribed and accomplished in 

accordance with documented instructions, procedures, or drawings. 

b. Measures to establish provisions that clearly delineate the sequence of actions to be 

accomplished in the preparation, review, approval, and control of instructions, 

procedures, and drawings. 

c. Measures to ensure instructions, procedures, and drawings specify the methods for 

complying with each of the applicable QA criteria. 

d. Measures to ensure instructions, procedures, and drawings include quantitative 

acceptance criteria (such as dimensions, tolerances, and operating limits) as well as 

qualitative acceptance criteria (such as workmanship samples) as verification that 

activities important to safety have been satisfactorily accomplished. 

e. Measures to ensure the QA organization reviews and concurs with the procedures, 

drawings, and specifications related to inspection plans, tests, calibrations, and special 

processes as well as any subsequent changes to these documents. 

14.5.6 Document Control 

The QAPD should define the applicant's proposed procedures for preparing, issuing, and 

revising documents that specify quality requirements or prescribe activities affecting quality. 

The following are examples of areas/items that may be addressed to support implementation of 

the quality criteria: 
 14-9 

a. The QAPD should identify all documents to be controlled under this subsection. As a 

minimum, this should include design specifications; design and fabrication drawings; 

procurement documents; QA manuals; design criteria documents; fabrication, 

inspection, and testing instructions; and test procedures. 

b. Measures to ensure establishment of procedures to control the review, approval, and 

issuance of documents and changes thereto before release to ensure that the 

documents are adequate and applicable quality requirements are stated. 

c. Measures to ensure establishment of provisions to identify individuals or groups 

responsible for reviewing, approving, and issuing documents and revisions thereto. 

d. Measures to ensure document revisions receive review and approval by the same 

organizations that performed the original review and approval or by other qualified 

responsible organizations designated by the applicant. 

e. Measures to ensure that approved changes be included in instructions, procedures, 

drawings, and other documents before the change is implemented. 

f. Measures to ensure the control of obsolete or superseded documents to prevent 

inadvertent use. 

g. Measures to ensure documents are available at the location where the activity is 

performed. 

h. Measures to ensure establishment of a master list (or equivalent) to identify the current 

revision number of instructions, procedures, specifications, drawings, and procurement 

documents. In addition, measures to ensure updating of the list and distribution of it to 

predetermined, responsible personnel to preclude use of superseded documents. 

14.5.7 Control of Purchased Material, Equipment, and Services 

The QAPD should define the applicant's proposed procedures for controlling purchased 

material, equipment, and services to ensure conformance with specified requirements. The 

following are examples of areas/items that may be addressed to support implementation of the 

quality criteria: 

a. Measures to ensure qualified personnel evaluate the supplier's capability to provide 

services and products of acceptable quality before the award of the procurement order 

or contract. In addition, measures to ensure QA and engineering groups participate in 

the evaluation of those suppliers providing critical items and services important to safety, 

and the applicant should define the responsibilities for each group's participation. 

b. Measures to ensure evaluation of suppliers on the basis of one or more of the following 

criteria: 

• The supplier's capability to comply with the elements of the QA criteria that are 

applicable to the type of material, equipment, or service being procured. 

• Review of previous records and performance of suppliers who have provided 

similar articles or services of the type being procured. 
 14-10 

• A survey of the supplier's facilities and QA program to assess the capability to 

supply a product that meets applicable design, manufacturing, and quality 

requirements. 

c. Measures to ensure documentation and filing of the results of supplier evaluations. 

d. Measures to ensure planning and performing adequate surveillance of suppliers during 

fabrication, inspection, testing, and shipment of materials, equipment, and components 

in accordance with written procedures to ensure conformance to the purchase order 

requirements. In addition the measures should ensure that the procedures provide the 

following information: 

• Instructions that specify the characteristics or processes to be witnessed, 

inspected or verified, and accepted; the method of surveillance and the extent of 

documentation required; and those responsible for implementing these 

instructions. 

• Procedures for audits and surveillance to ensure that the supplier complies with 

the quality requirements (surveillance should be performed for SSCs for which 

verification of procurement requirements cannot be determined upon receipt). 

e. Measures to ensure the supplier furnish the following records to the purchaser: 

• Documentation that identifies the purchased material or equipment and the 

specific procurement requirements (e.g., codes, standards, and specifications) 

met by the items. 

• Documentation that identifies any procurement requirements that have not been 

met and a description of any nonconformances designated “accept as is” or 

“repair.” 

f. Measures to describe the proposed procedures for reviewing and accepting these 

documents and, as a minimum, to ensure that this review and acceptance will be 

undertaken by a responsible QA individual. 

g. Measures to ensure the conduct periodic audits, independent inspections, or tests to 

ensure the validity of the suppliers' certificates of conformance. 

h. Measures to ensure the performance of a receiving inspection of the supplier-furnished 

material, equipment, and services to ensure fulfillment of the following criteria: 

• The material, component, or equipment should be properly identified in a manner 

that corresponds with the identification on the purchasing and receiving 

documentation. 

• Material, components, equipment, and acceptance records should be inspected 

and judged acceptable in accordance with predetermined inspection instructions 

before installation or use. 
 14-11 

• Inspection records or certificates of conformance attesting to the acceptance of 

material, components, and equipment should be available before installation or 

use. 

• Items accepted and released should be identified as to their inspection status 

before they are forwarded to a controlled storage area or released for installation 

or further work. 

i. Measures to assess the effectiveness of suppliers' quality controls at intervals consistent 

with the importance to safety, complexity, and quantity of the SSCs procured. 

14.5.8 Identification and Control of Materials, Parts, and Components 

The QAPD should define the applicant's proposed provisions for identifying and controlling 

materials, parts, and components to ensure that incorrect or defective SSCs are not used. The 

following are examples of areas/items that may be addressed to support implementation of the 

quality criteria: 

a. Measures to establish procedures to identify and control materials, parts, and 

components (including partially fabricated subassemblies). 

b. Measures to determine identification requirements during generation of specifications 

and design drawings. 

c. Measures to ensure that identification will be maintained either on the item or on records 

traceable to the item to preclude use of incorrect or defective items. 

d. Measures to ensure Identification of materials and parts of important-to-safety items are 

traceable to the appropriate documentation (such as drawings, specifications, purchase 

orders, manufacturing and inspection documents, deviation reports, and physical and 

chemical mill test reports). 

e. Measures to ensure the location and method of identification do not affect the fit, 

function, or quality of the item being identified. 

f. Measures to verify and document the correct identification of all materials, parts, and 

components before releasing them for fabrication, assembly, shipping, and installation. 

14.5.9 Control of Special Processes 

The QAPD should describe the controls that the applicant will establish to ensure the 

acceptability of special processes (such as welding, heat treatment, nondestructive testing, and 

chemical cleaning) and that the proposed controls are performed by qualified personnel using 

qualified procedures and equipment. The following are examples of areas/items that may be 

addressed to support implementation of the quality criteria: 

a. Measures to establish procedures to control special processes (such as welding, heat 

treating, nondestructive testing, and cleaning) for which direct inspection is generally 

impossible or disadvantageous. In addition, the applicant should provide a listing of 

these special processes. 
 14-12 

b. Measures to qualify procedures, equipment, and personnel connected with special 

processes in accordance with applicable codes, standards, and specifications. 

c. Measures to ensure qualified personnel perform special processes in accordance with 

written process sheets (or the equivalent) with recorded evidence of verification. 

d. Measures to establish, file, and keep current qualification records of procedures, 

equipment, and personnel associated with special processes. 

14.5.10 Licensee Inspection 

The QAPD should define the applicant's proposed provisions for inspection of activities affecting 

quality to verify conformance with instructions, procedures, and drawings. The following are 

examples of areas/items that may be addressed to support implementation of the quality 

criteria: 

a. Measures to establish, document, and conduct an inspection program that effectively 

verifies conformance of quality-affecting activities with requirements in accordance with 

written, controlled procedures. 

b. Measures to ensure inspection personnel are sufficiently independent from the 

individuals performing the activities being inspected. 

c. Measures to ensure inspection procedures, instructions, and check lists provide the 

following details: 

 • Identification of characteristics and activities to be inspected. 

• Identification of the individuals or groups responsible for performing the 

inspection operation. 

 • Acceptance and rejection criteria. 

 • A description of the method of inspection. 

• Procedures for recording evidence of completing and verifying a manufacturing, 

inspection, or test operation. 

• Identification of the recording inspector or data recorder and the results of the 

inspection operation. 

d. Measures to ensure the use of inspection procedures or instructions with the necessary 

drawings and specifications when performing inspection operations. 

e. Measures to qualify inspectors in accordance with applicable codes, standards, and 

company training programs and in addition keeping inspector's qualifications and 

certifications current. 

f. Measures to inspect modifications, repairs, and replacements in accordance with the 

original design and inspection requirements or acceptable alternatives. 
 14-13 

g. Measures to establish provisions that identify mandatory inspection hold points for 

witnessing by a designated inspector. 

h. Measures to identify the individuals or groups who will perform receiving and process 

verification inspections, and should demonstrate that these individuals or groups have 

sufficient independence and qualifications. 

I. Measures to establish provisions for indirect control by monitoring processing methods, 

equipment, and personnel if direct inspection is not possible. 

14.5.11 Test Control 

The QAPD should define the applicant's proposed provisions for tests to verify that SSCs 

conform to specified requirements and will perform satisfactorily in service. The following are 

examples of areas/items that may be addressed to support implementation of the quality 

criteria: 

a. Measures to establish, document, and conduct a test program to demonstrate that the 

item will perform satisfactorily in service in accordance with written, controlled 

procedures. 

b. Measures to ensure written test procedures incorporate or reference the following 

information: 

• Requirements and acceptance limits contained in applicable design and 

procurement documents. 

 • Instructions for performing the test. 

 • Test prerequisites. 

 • Mandatory inspection hold points. 

 • Acceptance and rejection criteria. 

 • Methods of documenting or recording test data results. 

c. Measures to ensure a qualified, responsible individual or group document test results 

and evaluate their acceptability. When practicable, the measures should ensure testing 

of the SSC occurs under conditions that will be present during normal and anticipated 

off-normal operations. 

14.5.12 Control of Measuring and Test Equipment 

The QAPD should define the applicant’s proposed provisions to ensure that tools, gauges, 

instruments, and other measuring and testing devices are properly identified, controlled, 

calibrated, and adjusted at specified intervals. The following are examples of areas/items that 

may be addressed to support implementation of the quality criteria: 

a. Measures to ensure documented procedures describe the calibration technique and 

frequency, maintenance, and control of all measuring and test equipment (instruments, 
 14-14 

tools, gauges, fixtures, reference and transfer standards, and nondestructive test 

equipment) that will be used in the measurement, inspection, and monitoring of SSCs 

that are important to safety. 

b. Measures to ensure measuring and test equipment are identified and traceable to the 

calibration test data. 

c. Measures to ensure the use of labels, tags, or documents for measuring and test 

equipment to indicate the date of the next scheduled calibration and to provide 

traceability to calibration test data. 

d. Measures to calibrate measuring and test instruments at specified intervals on the basis 

of the required accuracy, precision, purpose, degree of usage, stability characteristics, 

and other conditions that could affect the accuracy of the measurements. 

e. Measures to assess the validity of previous inspections when measuring and test 

equipment is found to be out of calibration. In addition, measures should also be 

provided to document the assessment and take control of the out of calibration 

equipment. 

f. Measures to document and maintain the complete status of all items under the 

calibration system. 

g. Measures to ensure reference and transfer standards are traceable to nationally 

recognized standards; where national standards do not exist, the applicant should 

establish provisions to document the basis for calibration. 

14.5.13 Handling, Storage, and Shipping Control 

The QAPD should define the applicant’s proposed provisions to control the handling, storage, 

shipping, cleaning, and preservation of SSCs in accordance with work and inspection 

instructions to prevent damage, loss, and deterioration. The following are examples of 

areas/items that may be addressed to support implementation of the quality criteria: 

a. Measures to establish and accomplish special handling, preservation, storage, cleaning, 

packaging, and shipping requirements in accordance with predetermined work and 

inspection instructions. 

b. Measures to control the cleaning, handling, storage, packaging, shipping, and 

preservation of materials, components, and systems in accordance with design and 

specification requirements to preclude damage, loss, or deterioration by environmental 

conditions (such as temperature or humidity). 

14.5.14 Inspection, Test, and Operating Status 

The QAPD should define the applicant’s proposed provisions to control the inspection, test, and 

operating status of SSCs to prevent inadvertent use or bypassing of inspections and tests. The 

following are examples of areas/items that may be addressed to support implementation of the 

quality criteria: 

a. Measures to know the inspection and test status of items throughout fabrication. 
 14-15 

b. Measures to establish procedures to control the application and removal of inspection 

and welding stamps and operating status indicators (such as tags, markings, labels, and 

stamps). 

c. Measures to ensure procedures under the cognizance of the QA organization controls 

the bypassing of required inspections, tests, and other critical operations. 

d. Measures to specify the organization responsible for documenting the status of 

nonconforming, inoperative, or malfunctioning SSCs and identifying the item to prevent 

inadvertent use. 

14.5.15 Nonconforming Materials, Parts, or Components 

The QAPD should define the applicant’s proposed provisions to control the use or disposition of 

nonconforming materials, parts, or components. The following are examples of areas/items that 

may be addressed to support implementation of the quality criteria: 

a. Measures to establish procedures to control the identification, documentation, tracking, 

segregation, review, disposition, and notification of affected organizations regarding 

nonconforming materials, parts, components, services, or activities. 

b. Measures to provide for adequate documentation to identify nonconforming items and 

describe the nonconformance, its disposition, and the related inspection requirements. 

The measures should also provide for adequate documentation and include signature 

approval of the disposition. 

c. Measures to establish provisions to identify those individuals or groups with the 

responsibility and authority for the disposition and closeout of nonconformance. 

d. Measures to ensure nonconforming items are segregated from acceptable items and 

identified as discrepant until properly dispositioned and closed out. 

e. Measures to verify the acceptability of reworked or repaired materials, parts, and SSCs 

by re-inspecting and retesting the item as originally inspected and tested or by using a 

method that is at least equal to the original inspection and testing method. In addition, 

the measures should provide for documentation of the relevant inspection, testing, 

rework, and repair procedures. 

f. Measures to ensure nonconformance reports designated “accept as is” or “repair” are 

made part of the inspection records and forwarded with the hardware to the customer for 

review and assessment. 

g. Measures to periodically analyze nonconformance reports to show quality trends and 

help identify root causes of nonconformance. Significant results should be reported to 

responsible management for review and assessment. 

14.5.16 Corrective Action 

The QAPD should define the applicant’s proposed provisions to ensure that conditions adverse 

to quality are promptly identified and corrected, and that measures are taken to preclude 
 14-16 

recurrence. The following are examples of areas/items that may be addressed to support 

implementation of the quality criteria: 

a. Measures to evaluate conditions adverse to quality (such as nonconformance, failures, 

malfunctions, deficiencies, deviations, and defective material and equipment) in 

accordance with established procedures to assess the need for corrective action. 

b. Measures to initiate corrective action to preclude recurrence of a condition identified as 

adverse to quality. 

c. Measures to conduct follow-up activities to verify proper implementation of corrective 

actions and close out the corrective action documentation in a timely manner. 

d. Measures to document significant conditions adverse to quality, as well as the root 

causes of the conditions, and the corrective actions taken to remedy the and preclude 

recurrence of the conditions. In addition, this information should be reported to 

cognizant levels of management for review and assessment. 

14.5.17 Quality Assurance Records 

The SAR should define the applicant’s proposed provisions for identifying, retaining, retrieving, 

and maintaining records that document evidence of the control of quality for activities and SSCs 

important to safety. The following are examples of areas/items that may be addressed to 

support implementation of the quality criteria: 

a. Measures to define the scope of the records program such that sufficient records will be 

maintained to provide documentary evidence of the quality of items and activities 

affecting quality. To minimize the retention of unnecessary records, the records program 

should list records to be retained by “type of data” rather than by record title. 

b. Measures to ensure that QA records include operating logs; results of reviews, 

inspections, tests, audits, and material analyses; monitoring of work performance; 

qualification of personnel, procedures, and equipment; and other documentation such as 

drawings, specifications, procurement documents, calibration procedures and reports, 

design review and peer review reports, nonconformance reports, and corrective action 

reports. 

c. Measures to ensure records are identified and retrievable. 

d. Measures to ensure requirements and responsibilities for record creation, transmittal, 

retention (such as duration, location, fire protection, and assigned responsibilities), and 

maintenance subsequent to completion of work are consistent with applicable codes, 

standards, and procurement documents. 

e. Measures to ensure inspection and test records contain the following information, where 

applicable: 

 • A description of the type of observation. 

 • The date and results of the inspection or test. 

 • Information related to conditions adverse to quality. 

 • Identification of the inspector or data recorder. 
 14-17 

 • Evidence as to the acceptability of the results. 

 • Action taken to resolve any noted discrepancies. 

f. Measures to ensure record storage facilities are constructed, located, and secured to 

prevent destruction of the records by fire, flood, theft, and deterioration by environmental 

conditions (such as temperature or humidity). In addition, the facilities are to be 

maintained by, or under the control of, the licensee throughout the life of the DSS or the 

individual product. 

14.5.18 Audits 

The QAPD should define the applicant’s proposed provisions for planning and scheduling audits 

to verify compliance with all aspects of the QA program, and to determine the effectiveness of 

the overall program. The following are examples of areas/items that may be addressed to 

support implementation of the quality criteria: 

a. Measures to perform audits in accordance with written procedures or checklists; 

qualified personnel tasked with performing these audits should not have direct 

responsibility for the achievement of quality in the areas being audited. 

b. Measures to ensure audit results are documented and reviewed with management 

having responsibility in the area audited. 

c. Measures to establish provisions for responsible management to undertake appropriate 

corrective action as a follow-up to audit reports. In addition, the measures should 

ensure auditing organizations schedule and conduct appropriate follow-up to ensure that 

the corrective action is effectively accomplished. 

d. Measures to perform both technical and QA programmatic audits to achieve the 

following objectives: 

• Provide a comprehensive independent verification and evaluation of procedures 

and activities affecting quality. 

 • Verify and evaluate suppliers’ QA programs, procedures, and activities. 

e. Measures to ensure audits are led by appropriately qualified and certified audit 

personnel from the QA organization. The measures should also ensure that the audit 

team membership include personnel (not necessarily QA organization personnel) having 

technical expertise in the areas being audited. 

f. Measures to schedule regular audits on the basis of the status and importance to safety 

of the activities being audited. The measures should also address that audits be 

initiated early enough to ensure effective QA during design, procurement, and 

contracting activities. 

g. Measures to analyze and trend audit deficiency data as well as ensuring resultant 

reports, indicating quality trends and the effectiveness of the QA program, should be 

given to management for review, assessment, corrective action, and follow-up. 
 14-18 

h. Measures to ensure that audits objectively assess the effectiveness and proper 

implementation of the QA program and should address the technical adequacy of the 

activities being conducted. 

I. Measures to establish provisions requiring the performance of audits in all areas to 

which the requirements of the QA program apply. 

14.6 Evaluation Findings 

If the reviewer determines that the applicant’s QAPD does not adequately address the Part 72 

requirements, a request for additional information (RAI) must be prepared and submitted to the 

Project Manager to be forwarded to the applicant for resolution and response to the NRC. If the 

reviewer concludes that information provided with the application, along with additional 

information provided in response to NRC RAI(s), shows that the QA program description meets 

the acceptance requirements referenced in Section 14.4, findings of the following type should 

be included in the staff’s SER or in a letter to the applicant, if the applicant’s QA program 

description was submitted separate from a SAR. 

(finding numbering is for convenience in referencing within the FSRP and SER): 

F14.1 Based upon a review and evaluation of the QA program description contained in the 

Safety Analysis Report or applicant’s submittal (identified by date and any other pertinent 

identifiers) for a DSS, the staff concludes that: 

• The licensee’s description of the QA program indicates requirements, 

procedures, and controls that, when properly implemented, should comply with 

the requirements of 10 CFR 72, Subpart G. 

• The licensee’s description of the QA program covers activities affecting SSCs 

important to safety as identified in the Safety Analysis Report. 

• The licensee’s description of the QA program describes organizations and 

persons performing QA functions indicating that sufficient independence and 

authority should exist to perform their functions without undue influence from 

those directly responsible for costs and schedules. 

• The licensee’s description of the QA program is in compliance with applicable 

NRC regulations and industry standards, and the acceptance of the QA program 

description by NRC allows implementation of the associated QA program for the 

(specify: design, fabrication and construction, operation, decommissioning) 

phases of the installation’s life cycle. 
 A-1 

APPENDIX A CONSOLIDATED REFERENCES 

A.1 U.S. Nuclear Regulatory Commission (NRC) Documents Cited

A.1.1 U.S. Code of Federal Regulations (CFR), Title 10, “Energy”

Part 2, “Rules of Practice for Domestic Licensing Proceedings and Issuance of Orders,” 

August 15, 1991. 

Part 20, “Standards for Protection Against Radiation,” September 11, 1988. 

Part 50, “Domestic Licensing of Production and Utilization Facilities,” August 15, 1991. 

Part 71, “Packaging and Transportation of Radioactive Material,” Appendix H, Quality 

Assurance, September 28, 1995. 

Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, HighLevel Radioactive Waste, and Reactor-Related Greater Than Class C Waste,” January 1, 2001. 

Part 73, “Physical Protection of Plants and Materials,” December 28, 1973. 

Part 100, “Reactor Site Criteria,” January 10, 1997. 

Part 961, “Standard Contract for Disposal of Spent Nuclear Fuel and/or High Level Radioactive 

Waste,” April 18, 1983. 

A.1.2 Regulatory Guides (RG)

RG 1.25, “Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel 

Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water 

Reactors,” March 1972. 

RG 1.26, “Quality Group Classification and Standards for Water-, Steam-, and RadioactiveWaste-Containing Components of Nuclear Power Plants,” Revision 4, March 2007, 

ML070290283. 

RG 1.29, “Seismic Design Classification,” Revision 4, March 2007, ML070310052. 

RG 1.33, “Quality Assurance Program Requirements (Operation),” Revision 3, February 1978, 

ML0037399. 

RG 1.59, “Design Basis Floods for Nuclear Power Plants,” Revision 2, August 1977 with Errata 

of 7/30/1980, ML003740388. 

RG 1.60, “Design Response Spectra for Seismic Design of Nuclear Power Plants,” Revision 1, 

December 1973. ML003740207. 

RG 1.61, “Damping Values for Seismic Design of Nuclear Power Plants,” Revision 1, March 

2007, ML070260029. 
 A-2 

RG 1.76, “Design Basis Tornado and Tornado Missiles for Nuclear Power Plants,” Revision 1, 

March 2007, ML070360253. 

RG 1.86, “Termination of Operating Licenses for Nuclear Reactors,” June 1974, ML003740243. 

RG 1.92, “Combining Modal Responses and Spatial Components in Seismic Response 

Analysis,” Revision 2, July 2006, ML053250475. 

RG 1.102, “Flood Protection for Nuclear Power Plants,” Revision 1, September 1976, 

ML003740308. 

RG 1.109, “Calculations of Annual Doses to Man from Routine Releases of Reactor Effluents for 

the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,” Revision 1, October 

1977, ML003740384. 

RG 1.117, “Tornado Design Classification,” Revision 1, April 1978, ML003739346. 

RG 1.136, “Design Limits, Loading Combinations, Materials, Construction, and Testing of 

Concrete Containments,” Revision 3, March 2007, ML070310045. 

RG 1.142, “Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor 

Vessels and Containments),” Revision 2, November 30, 2001, ML013100274. 

RG 1.143, “Design Guidance for Radioactive Waste Management Systems, Structures, and 

Components Installed in Light-Water-Cooled Nuclear Power Plants,” Revision 2, November 

2001, ML013100305. 

RG 1.145, “Atmospheric Dispersion Models for Potential Accident Consequence Assessments 

at Nuclear Power Plants,” February 1989, ML003740205. 

RG 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at 

Nuclear Power Plants,” Revision 0, July 2000, ML003716792. 

RG 1.193, “ASME Code Cases Not Approved for Use,” Revision 2, October 2007, 

ML072470294. 

RG 3.60, “Design of an Independent Spent Fuel Storage Installation (Dry Storage),” 

March 1997, ML003739501. 

RG 3.61, “Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel 

Dry Storage Cask,” February 1989, ML003739511. 

RG 7.11, “Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask 

Containment Vessels with a Maximum Wall Thickness of 4 Inches,” Revision 0, June 1991, 

ML003739413. 

RG 7.12, “Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask 

Containment Vessels with a Wall Thickness Greater Than 4 Inches, But Not Exceeding 

12 Inches,” Revision 0, June 1991, ML003739424. 
 A-3 

RG 8.5, “Criticality and Other Interior Evaluation Signals, Revision 1, March 1981, 

ML003739454. 

RG 8.8, “Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear 

Power Stations Will Be as Low as Reasonably Achievable,” June 2001. 

RG 8.10, “Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as is 

Reasonably Achievable,” September 1975, ML003739563. 

RG 8.25, “Air Sampling in the Workplace,” Revision 1, June 1992, ML003739616. 

RG 8.34, “Monitoring Criteria and Methods to Calculate Occupational Radiation Doses,” 

July 1992, ML003739502. 

RG 8.36, “Radiation Dose to the Embryo/Fetus,” July 1992, ML003739548. 

A.1.3 NUREG

NUREG-0612, “Control of Heavy Loads at Power Plants,” July 1980. 

NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear 

Power Plants,” March, 2007. 

NUREG-1567, “Standard Review Plan for Spent Fuel Dry Storage Facilities,” March 2000. 

NUREG-1571, “Information Handbook on Independent Spent Fuel Storage Installations,” 

Raddatz, M.G. and Waters, M.D., December 1995. 

NUREG-1614, “Strategic Plan, FY2004 - FY2009,” Volume 3, August 2004. 

NUREG-1727, “NMSS Decommissioning Standard Review Plan,” September 2000. 

NUREG-1745, “Standard Format and Content for Technical Specifications for 10 CFR Part 72 

Cask Certificates of Compliance,” June 2001. 

NUREG-1864, “A Pilot Probabilistic Risk Assessment of a Dry Cask Storage System at a 

Nuclear Power Plant,” March 2007 

A.1.4 NUREG/CR

NUREG/CR-1815, “Recommendations for Protecting Against Failure by Brittle Fracture in 

Ferritic Steel Shipping Containers Up to Four Inches Thick,” LLNL, June 1981. 

NUREG/CR-3826, “Recommendations for Protecting Against Failure by Brittle Fracture in 

Ferritic Steel Shipping Containers Greater than Four Inches Thick,” LLNL, July 1994. 

NUREG/CR-4554, “SCANS (Shipping Cask Analysis System): A Microcomputer Based Analysis 

System for Shipping Cask Design Review,” LLNL, March 1998. 

NUREG/CR-4775, “Guide for Preparing Operating Procedures for Shipping Packages,” 

UCID-20820, July 1988. 
 A-4 

NUREG/CR-5502, “Engineering Drawings for 10 CFR Part 71 Package Approval,” Lawrence 

Livermore National Laboratory (LLNL), May 1998. 

NUREG/CR-6007, “Stress Analysis of Closure Bolts for Shipping Casks,” Kaiser Engineering, 

January 1993. 

NUREG/CR-6242, “CASKS (Computer Analysis of Storage Casks): A Microcomputer-Based 

Analysis System for Storage Cask Design Review,” Lawrence Livermore National Laboratory 

(LLNL), February 1995. 

NUREG/CR-6322, “Buckling Analysis of Spent Fuel Basket”, UCRL-ID-119697, LLNL, 

May 1995. 

NUREG/CR-6328, “Adequacy of the 123-Group Cross-Section Library for Criticality Analyses of 

Water-Moderated Uranium Systems,” ORNL/TM-12970, ORNL, August 1995. 

NUREG/CR-6361, “Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation 

and Storage Packages.” ORNL/TM-13211, U.S. Nuclear Regulatory Commission (NRC), ORNL, 

March 1997. 

NUREG/CR-6407, “Classification of Transportation Packaging and Dry Spent Fuel Storage 

System Components According to Importance to Safety,” INEL-95/0551, Idaho National 

Engineering Laboratory (INEL), February 1996. 

NUREG/CR-6487, “Containment Analysis for Type B Packages Used to Transport Various 

Contents,” (LLNL), November 1996. 

NUREG/CR-6608, “Summary and Evaluation of Low-Velocity Impact Tests of Solid Steel Billet 

onto Concrete Pad,” Lawrence Livermore National Laboratory, February 1998. 

NUREG/CR-6700, “Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms 

Related to Transport and Interim Storage of High-Burnup LWR Fuel,” ORNL/TM-2000/284, 

ORNL, January 2001. 

NUREG/CR-6701, “Review of Technical Issues Related to Predicting Isotopic Compositions and 

Source Terms for High-Burnup LWR Fuel,” ORNL/TM-2000/277, ORNL, January 2001. 

NUREG/CR-6716, “Recommendations on Fuel Parameters for Standard Technical 

Specifications for Spent Fuel Storage Casks,” ORNL/TM-2000/385, Oak Ridge National 

Laboratory (ORNL), March 2001. 

NUREG/CR-6759, “Parametric Study of the Effect of Control Rods for PWR Burnup Credit,” 

ORNL/TM-2001/69, ORNL, February 2002. 

NUREG/CR-6760, “Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit,” 

ORNL/TM-2000/321, ORNL, March 2002. 

NUREG/CR-6761, “Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup 

Credit,” ORNL/TM-2000/373, ORNL, March 2002. 
 A-5 

NUREG/CR-6798, “Isotopic Analysis of High-Burnup PWR Spent Fuel Samples From the 

Takahama-3 Reactor,” ORNL/TM-2001/259, ORNL, January 2003. 

NUREG/CR-6801, “Recommendations for Addressing Axial Burnup in PWR Burnup Credit 

Analyses,” March 2003. 

NUREG/CR-6802, “Recommendations for Shielding Evaluations for Transport & Storage 

Packages,” May 2003. 

NUREG/CR-6835, “Effects of Fuel Failure on Criticality Safety and Radiation Dose for Spent 

Fuel Casks,” ORNL/TM-2002/255, ORNL, September 2003. 

A.1.5 Other NRC Publications

“Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety 

Related Equipment,” NRC Bulletin 96-02, April 11, 1996. 

“Chemical, Galvanic, or Other Reactions in Spent Fuel Storage and Transportation Casks,” 

NRC Bulletin 96-04, July 1996. 

Confirmatory Action Letter 97-7-001, July 22, 1998. 

Information Notice No. 91-26, “Potential Nonconservative Errors in the Working Format HansenRoach Cross Section Set Provided with the KENO and SCALE Codes,” April 15, 1991. 

Tang, David T., et al., “NRC Staff Technical Approach for Spent Fuel Storage Cask Drop and 

Tipover Accident Analysis,” Spent Fuel Project Office, 1997. 

A.2 Codes, Standards, and Specifications

American Concrete Institute (ACI), “Code Requirements for Nuclear Safety-Related Concrete 

Structures and Commentary,” ACI 349-06/349R-06, 2006. 

– – – “Building Code Requirements for Structural Plain Concrete and Commentary,” 

ACI 318-05/310R-05 with Errata of 0/11/05, 2005. 

– – – “Building Code Requirements for Masonry Structures, and Commentary,” ACI 530-05, 

2005. 

American Society of Mechanical Engineers (ASME), “Cases of ASME Boiler and Pressure 

Vessel Code,” Code Case N-595-4, 2004. 

American Society of Mechanical Engineers, Boiler and Pressure Vessel (B&PV) Code, 

“Specification for Welding Rods, Electrodes and Filler Metals,” Section II, “Materials” - Part C, 

2001. 

– – – Section III, “Rules for Construction of Nuclear Facility Components,” 2007. 

Division 1 - General Requirements for Division 1 and Division 2; Subsection NB through 

NH, and Appendices. 

 Division 2 - Code for Concrete Containment (Also known as ACI 359-07). 
 A-6 

Division 3 - Containment for Transportation and Storage of Spent Nuclear Fuel and High 

Level Radioactive Material and Waste. 

– – – ASME B&PV Code, Section V, “Nondestructive Examination Specifications and 

Procedures.” 

– – – ASME B&PV Code, Section VIII, Division 3, “Alternative Rules for the Construction of High 

Pressure Vessels,” 2001. 

– – – ASME B&PV Code, Section IX, “Qualification Standard for Welding and Brazing 

Procedures, Welders, Brazers, and Welding and Brazing Operators.“ 

– – – ASME B&PV Code, Section XI, “Rules for Inservice Inspection of Nuclear Power Plant 

Components,” 2001. 

American Institute of Steel Construction (AISC), “Code of Standard Practice for Steel Buildings 

and Bridges,” March 2005. 

– – – “Specification for Structural Steel Buildings,” March 2005. 

ANSI/American Nuclear Society (ANS), “Design Criteria for an Independent Spent Fuel Storage 

Installation (Dry Storage Type),” ANSI/ANS 57.9-1992-R2000, 2000. 

– – – “Neutron and Gamma-Ray Flux-to-Dose Conversion Factors,” ANSI/ANS-6.1.1, 1977. 

– – – “Neutron and Gamma-Ray Flux-to-Dose Conversion Factors,” ANSI/ANS-6.1.1, 1991. 

– – – “Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors,” 

ANSI/ANS-8.1, 1998. 

– – – “Administrative Controls and Quality Assurance for the Operational Phase of Nuclear 

Power Plants,” ANSI/ANS 3.2. 

ANSI, Institute for Nuclear Materials Management, “American National Standard for Leakage 

Tests on Packages for Shipment of Radioactive Materials,” ANSI N14.5, 1997. 

– – – “American National Standards for Radioactive Materials-Special Lifting Devices for 

Shipping Containers Weighing 10,000 Pounds (4500 Kilograms) or More,” ANSI N14.6-1986, 

1986. 

– – – “Characterizing Damaged Spent Nuclear Fuel for the Purpose of Storage and Transport,” 

ANSI N14.33-2005. 

ANSI/American Nuclear Society (ANS), “Requirements for Collection, Storage, and 

Maintenance of Quality Assurance for Nuclear Power Plants,” ANSI/ASME N45.2.9-1979. 

ANSI/ANS, “Nuclear Facilities – Steel Safety Related Structures for Design Fabrication and 

Erection,” N690. 

ANSI/ASME B16.34, “Valves Flanged, Threaded and Welding End.” 
 A-7 

ANSI/ASME B31.1, “Power Piping.” 

ANSI/ASME B96.1, “Specification for Welded Aluminum-Alloy Field-Erected Storage Tanks.” 

ANSI/ASME NQA-1, “Quality Assurance Program for Nuclear Facilities.” 

ANSI/ASME NQA-2, “Quality Assurance Requirements for Nuclear Facilities.” 

American Petroleum Institute (API), “Recommended Rules for Design and Construction of Large 

Welded, Low-Pressure Storage Tanks,” API 620, February 2002. 

American Society of Civil Engineers (ASCE), “Minimum Design Loads for Buildings and Other 

Structures,” ASCE 7-05, 2005. 

– – – “Seismic Analysis of Safety-Related Nuclear Structures,” ASCE 4-98, 2002. 

American Society for Testing and Materials International (ASTM), “Draft 17-Guide for Evaluation 

of Materials Used in Extended Service of Interim Spent Nuclear Fuel Dry Storage Systems,” 

October 2002. 

– – – “Standard Practice for Prediction of the Long-Term Behavior of Waste Package Materials 

Including Waste Forms Used in the Geologic Disposal of High-Level Nuclear Waste,” C1174-97, 

2003. 

– – – “Standard Practice for Qualification and Acceptance of Boron Based Metal Neutron 

Absorbers for Nuclear Criticality Control for Dry Storage Systems and Transportation 

Packaging,” C1671. 

– – – “Standard Test Method for Dynamic Tear Testing of Metallic Materials,” ASTM E604-83, 

2002. 

– – – “Standard Test Method of Conducting Drop-Weight Test to Determine Nil-Ductility 

Transition Temperature of Ferritic Steels,” ASTM E208-95a, 2000. 

– – – “Standard Specification for Concrete Aggregates,” C 33, 2002. 

American Water Works Association (AWWA), “Welded Steel Tanks for Water Storage,” 

AWWA D100. 

American Welding Society (AWS), “Standard Symbols for Welding, Brazing, and Nondestructive 

Examination,” AWS A2.4 (Latest Edition). 

– – – “Structural Welding Code-Steel,” AWS D1.1/D1.1M-2002, 2002. 

International Commission on Radiological Protection (ICRP), “Statement from the 1980 Meeting 

of the ICRO,” ICRP Publication 26, Pergammon Press, New York, New York, 1980. 

International Conference Council (ICC), “International Building Code (IBC),” 2006. 

American Society for Nondestructive Testing (SNT), “Personnel Qualification and Certification in 

Nondestructive Testing,” SNT-TC-1A. 
 A-8 

A.3 Other Government Agencies

Environmental Protection Agency, “Manual of Protective Action Guides and Protective Actions 

for Nuclear Incidents,” EPA 410R-92-001. 

– – – “Manual of Protective Action Guides and Protective Actions for Nuclear Incidents,” 

EPA 410-R-92-001, May 1992. 

– – – “External Exposure to Radionuclides in Air, Water, and Soil,” EPA Guidance Report 

No. 12, 1993. 

Newman, L.W., “The Hot Cell Examination of Oconee Fuel Rods After Five Cycles of 

Irradiation,” DOE/ET/34212-50, U.S. Department of Energy (DOE), 1986. 

U.S. Department of Energy (DOE), “Criticality Safety Good Practices Program Guide for DOE 

Nonreactor Nuclear Facilities,” DOE G 421.1-1, August 25, 1999. 

Nuclear Science Committee, Nuclear Energy Agency, “International Handbook of Evaluated 

Criticality Safety Benchmark Experiments,” NEA/NSC/DOC(95)03, September 2003. (This 

document is updated and published annually in CD-ROM format). 

A.4 Technical Reports

AEA Technology, “MONK - A Monte Carlo Program for Nuclear Criticality Safety and Reactor 

Physics Analyses, User Guide for Version 8,” ANSWERS/MONK(98)6, June 1991. Issued 

through the ANSWERS Software Service. 

ANSYS, Inc., “ANSYS Basic Analysis Procedures Guide,” Fourth Edition, ANSYS Release 5.6, 

November 1999. 

Bechtel, “Commercial Spent Nuclear Fuel Handling in Air Study,” 000-30R-MGR0-00700- 

000000, March 2005. 

Beyer, C.E., Letter from C.E. Beyer, Pacific Northwest National Laboratory, to K. Gruss, 2001. 

Boase, D.G. and T.T. Vandergraaf, “The Canadian Spent Fuel Storage Canister: Some 

Materials Aspects,” Nucl. Techol., 32, 60, (1977). 

Bjorkman & Moore, Influence of ISFSI Design Parameters on the Seismic Response of Dry 

Storage Casks, 2001. 

Bjorkman, et al., Seismic Analysis of Plant Hatch ISFSI Pad and Stability Assessment of Dry 

Casks, 2000. 

Broadhead, B.L., et al., “Evaluating of Shielding Analysis Methods in Spent Fuel Cask 

Environments,” EPRI TR-104329, Electric Power Research Institute (EPRI), Palo Alto, 

California, May 1995. 

Cacciapouti, R.J., and S. Van Volkinburg. “Axial Burnup Profile Database for Pressurized Water 

Reactors.” YAEC-1937. May 1997. Available as Data Package DLC-201 from the Radiation 
 A-9 

Safety Information Computational Center at Oak Ridge National Laboratory (ORNL). 

http://www-rsicc.ornl.gov/ORDER.html. 

Cappelaere, R. Limon, T. Bredel, P. Herter, D. Gilbon, S. Allegre, P. Bouffioux and J.P. 

Mardon. “Long Term Behaviour of the Spent Fuel Cladding in Dry Storage Conditions.” 

8th International Conference on Radioactive Waste Management and Environmental 

Remediation. October 2001. Bruges, Belgium. 

Chung, H.M. and T.F. Kassner. “Cladding Metallurgy and Fracture Behavior During 

Reactivity-Initiated Accidents at High Burnup.” Proceedings of the International Topical 

Meeting on Light Water Reactor Fuel Performance. American Nuclear Society. 

March 2-6, 1997. Portland, Oregon. 1997. 

Chung, H.M. “Fundamental Metallurgical Aspects of Axial Splitting in Zircaloy Cladding.” 

 Proceedings of the International Topical Meeting on Light Water Reactor Fuel Performance. 

American Nuclear Society. April 10-13, 2000. Park City, UT. 2000. 

Chun, R., Witte, M., and Schartz, M., “Dynamic Impact Effects on Spent Fuel Assemblies,” 

UCID-21246, LLNL, October 20, 1987. 

Cottrell, W.B., and Savolainen, A.W., “U.S. Reactor Containment Technology,” ORNL-NSIC-5, 

Volume 1, Chapter 6, ORNL, August 1965. 

Cunningham, M.E., E.R. Gilbert, A.B. Johnson, and M.A. McKinnon, “Evaluation of Expected 

Behavior of LWR Stainless Steel-Clad Fuel in Long-Term Dry Storage,” EPRI TR-106440, 

April 1996. 

DeHart, M.D. and O.W. Hermann, “An Extension of the Validation of SCALE (SAS2H) Isotopic 

Prediction for PWR Spent Fuel,” ORNL/TM-13317, ORNL, September 1996. 

Eckerman, K.F. and J.C. Ryman, “External Exposure to Radionuclides in Air, Water, and Soil,” 

Federal Guidance Report No. 12, EPA 402-R-93-081, ORNL, September 1993. 

Einziger, R.E., et al., “Examination of Spent Fuel Rods After 15 Years in Dry Storage,” Argonne 

National Laboratory (ANL), 2002. 

Einziger, R. E., et al., “High Temperature Postirradiation Materials Performance of Spent 

Pressurized Water Reactor Fuel Rods Under Dry Storage Conditions,” Nuclear Technology, 

v. 57, p. 65, 1982. 

Einziger, R.E. and R. Kohli, “Low Temperature Rupture Behavior of Zircaloy-Clad Pressurized 

Water Reactor Spent Fuel Rods under Dry Storage Conditions,” Nuclear Technology, v. 67, 

p. 107, 1984. 

Einziger, R.E. and J.A. Cook, “LWR Spent Fuel Dry Storage Behavior at 229°C,” HEDLTME 

84-17, NUREG/CR-3708, Hanford Engineering Development Laboratory (Aug 1984). 

Einziger, R.E. and R.V. Strain, “Oxidation of Spent Fuel at Between 250° and 360°C,” EPRI 

Report NP-4524, 1986. 
 A-10 

Einziger, R.E., L.E. Thomas, H.V. Buchanan, and R.B. Stout, “Oxidation of Spent Fuel in Air 

at 175 to 195°C,” J Nucl. Mater., 190, p53., (1992). 

Federal Registry (FR), “List of Approved Spent Fuel Storage Casks: Holtec HI–STORM 100 

Addition,” Vol. 65, No. 84, pg. 25241, May 1, 2000. 

Ferry, C, et al. - Synthesis on the Spent Fuel Long Term Evolution, Rapport 

CEA-R6084, (2005). 

Fontana, M.G. and N.D. Greene, Corrosion Engineering, McGraw Hill, 1978. 

Gao, J., “Modeling of Neutron Attenuation Properties of Boron-Aluminum Shielding Materials,” 

Masters Dissertation, University of Virginia, August 1997. 

Garde, A.M., et al., “Effects of Hydride Precipitate Localization and Neutron Fluence on the 

Ductility of Irradiated Zircaloy-4,” Zirconium in the Nuclear Industry: Eleventh International 

Symposium, ASTM STP 1295, American Society for Testing and Materials (ASTM), 1996. 

Goll, W., et al., “Short-Term Creep and Rupture Tests on High Burnup Fuel Rod Cladding,” 

Journal of Nuclear Materials,” v. 289, p. 247, 2001. 

Hanson, B.D., 1998, “The Burnup Dependence of Light Water Reactor Spent Fuel Oxidation,” 

PNNL-11929, Richland, Washington, Pacific Northwest National Laboratory. TIC: 238459. 

Hermann, O.W. and M.D. DeHart, “Validation of SCALE (SAS2H) Isotopic Predictions for BWR 

Spent Fuel,” ORNL/TM-13315, ORNL, September 1998. 

Hoerner, S.F., Fluid-Dynamics Drag, Hoerner Fluid Dynamics, 1965. 

Johnson, A.B., et al., “Exposure of Breached BWR Fuel Rods at 325°C to Air and Argon,” 

Proc. NRC Workshop on Spent Fuel/Cladding Reaction During Dry Storage, 

Gaithersburg, Maryland, Aug 1983, NUREG/CR-0049, D. REISENWEAVER, 

Ed., S. Nuclear Regulatory Commission (1984). 

Kammenzind, B.F., et al., “The Long-Range Migration of Hydrogen Through Zircaloy in 

Response to Tensile and Compressive Stress Gradients,” Zirconium in the Nuclear Industry: 

Twelfth International Symposium, ASTM STP 1354, G.P. Sabol and G.D. Moan, Eds., 

American Society for Testing and Materials, pp. 196-233, 2000. 

Kennedy, R.P., Review of Procedures for the Analysis and Design of Concrete Structures to 

Resist Missile Impact Effects, Holmes and Narver, Inc., September 1975. 

Kese, K., “Hydride Re-Orientation in Zircaloy and its Effect on the Tensile Properties,” SKI 

Report 98:32, 1998. 

Knoll, R.W., et al., “Evaluation of Cover Gas Impurities and Their Effects on the Dry Storage of 

LWR Spent Fuel,” PNL-6365, DE88 003983, PNNL, November 1987. 

Lloyd, W.R., “Determination and Application of Bias Values in the Criticality Evaluation of 

Storage Cask Designs,” UCID-21830, LLNL, January 1990. 
 A-11 

Manteufel, R.D. and Todreas, N.E., “Effective Thermal Conductivity and Edge Configuration 

Model for Spent Fuel Assembly,” Nuclear Technology, Vol. 105, pp. 421-440, March 1994. 

Machiels, “Regulatory Applications Lessons Learned -- Industry Perspective.” NEI Dry 

Storage Information Forum. Naples, FL. May 15-16, 2002. 

MCNP5, “MCNP – A General Monte Carlo N-Particle Transport Code, Version 5; Volume II: 

User’s Guide,” LA-CP-03-0245, Los Alamos National Laboratory, April 2003. 

Nakamura, J., T. Otomo, T. Kikuchi, and S. Kawasaki, “Oxidation of Fuel Rods under 

Dry Storage Condition,” J Nuc. Sci. Tech., 32, [4], p321, (April 1995). 

National Association of Corrosion Engineers (NACE), Corrosion Data Survey, 1985. 

Novak, J., and I.J. Hastings, “Post-Irradiation Behavior of Defected UO2 in Air at 220250°C,” 

Proc. NRC Workshop on Spent Fuel/Cladding Reaction During Dry Storage, Gaithersburg, 

Maryland, Aug. 1983, NUREG/CR-0049, D. REISENWEAVER, Ed., S. Nuclear 

Regulatory Commission (1984). 

NRC. Subject: Transmittal of “Update of CSFM Methodology for Determining Temperature 

Limits for Spent Fuel Dry Storage in Inert Gas,” November 27, 2001. 

NRC Inspection Manual, Inspection Procedure 60851, “Design Control for ISFSI Components,” 

ML0037287650. 

Oak Ridge National Laboratory, “SCALE: A Modular Code System for Performing Standardized 

Computer Analyses for Licensing Evaluation,” ORNL/TM-2005/39, Version 5, Vols. I-III, 

April 2005. Available from Radiation Safety Information Computational Center at Oak Ridge 

National Laboratory as CCC-725. 

Pacific Northwest Laboratory (PNL), “Evaluation of Cover Gas Impurities and Their Effects on 

the Dry Storage of LWR Spent Fuel,” PNL-6365, November 1987. 

Parks, C.V., et al., “Assessment of Shielding Analysis Methods, Codes, and Data for Spent Fuel 

Transport/Storage Applications,” ORNL/CSD/TM-246, ORNL, July 1988. 

Rashid, Y.R. and R.S. Dunham, “Creep Modeling and Analysis Methodology for Spent Fuel in 

Dry Storage,” TR-1003135, EPRI, 2000. 

Rashid, Y.R., et al., “Creep as the Limiting Mechanism for Spent Fuel Dry Storage-Progress 

Report,” EPRI TR-1001207, EPRI, 2000. 

Roark, R.J., Formulas for Stress and Strain, McGraw Hill, 1965. 

Sandoval, R.P., et al., “Estimate of CRUD Contribution to Shipping Cask Containment 

Requirements,” SAND88-1358, TTC-0811, UC-71, SNL, January 1991. 

Stokley, J.R., and D.H. Williamson, “Structural Integrity of Spent Nuclear Fuel Storage Casks 

Subjected to Drop,” Nuclear Technology, Volume 114, Number 1, April 1996. 
 A-12 

TRW Environmental Safety Systems, Inc. (TRW), “DOE Characteristics Database, User Manual 

for the CDB-R,” November 16, 1992. 

Uhlig, H.H., Corrosion and Corrosion Control, Wiley & Sons, Inc., 1985. 

Wilson, D.W., et al., “Creep-Rupture Testing of Aluminum Alloys to 100,000 Hours, First 

Progress Report,” Prepared for the Metal Properties Council, New York, November, 1969. 

A.5 Correspondence

Beyer, C.E., PNNL, letter to K. Gruss, NRC, November 27, 2001, Subject: Transmittal of 

“Update of CSFM Methodology for Determining Temperature Limits for Spent Fuel Dry Storage 

Inert Gas,” November 27, 2001. 

Hendricks, L., Nuclear Energy Institute (NEI), letter to M.W. Hodges, NRC, Subject: Transmittal 

of Responses to the NRC Request for Additional Information on Storage of High Burnup Fuel, 

August 16, 2001. 

NRC Confirmatory Action Letter 97-7-001, 1998 (ADAMS ML060620420). 

Tsai, H.C. letter to K. Gruss, NRC, Subject: “A Recent Result on Thermal Creep of Surry 

Cladding after 15-y Dry Cask Storage,” ANL, July 11, 2002. 

Transnuclear (TN) Standardized NUHOMS Amendment 10 RAI Response, Docket No. 72-1004, 

November 7, 2007. 

A.6 Conference Proceedings

Cappelaere, C., R. Limon, T. Bredel, P. Herter, D. Gilbon, S. Allegre, P. Bouffioux and J.P. 

Mardon. 2001, “Long Term Behavior of the Spent Fuel Cladding in Dry Storage Conditions,” 8th 

International Conference on Radioactive Waste Management and Environmental Remediation, 

Bruges, Belgium, October 2001. 

Chung, H.M. and T.F. Kassner, “Cladding Metallurgy and Fracture Behavior During ReactivityInitiated Accidents at High Burnup,” Proceedings of the International Topical Meeting on Light 

Water Reactor Fuel Performance. American Nuclear Society, Portland, OR, March 2-6, 1997. 

Chung, H.M., “Fundamental Metallurgical Aspects of Axial Splitting in Zircaloy Cladding,” Park 

City, Utah, Proceedings of the International Topical Meeting on Light Water Reactor Fuel 

Performance, American Nuclear Society, Park City, Utah, April 10-13, 2000. 

Machiels, A., “Regulatory Applications Lessons Learned – Industry Perspective,” NEI Dry 

Storage Information Forum, Naples, Florida, May 15-16, 2002. 
 B-1 

APPENDIX B PROCESS FOR PRIORITIZING THE STANDARD REVIEW PLAN 

FOR DRY STORAGE SYSTEMS

B.1 Introduction 

The purpose of this appendix is to describe the process used for prioritizing the review 

procedures contained in this NUREG. The application of this process, which is based upon 

determining relative importance, has resulted in assigning priorities of HIGH, MEDIUM or LOW 

to each of the review procedures in the SRPs. These priorities are intended to help focus staff 

review resources on those review procedures which are considered to be the most effective and 

important to worker and public safety. They are not, however, intended to relieve applicants of 

responsibility to comply with all requirements associated with dry cask storage licensing. 

In 1995 the Commission issued a policy statement on the use of probabilistic risk assessment 

methods in all regulatory activities (60 FRN 42622, dated August 16, 1995). This policy 

statement has led to the development and application of "risk-informed" approaches in various 

regulatory areas. Specifically, a "risk-informed" approach represents a philosophy where risk 

insights are considered together with other factors to establish requirements that better focus 

licensee and regulatory attention on design and operational issues commensurate with their 

importance to safety. In general, "Risk-informed" approaches lie between "risk-based" and 

purely deterministic approaches, and are intended to: 

• Allow consideration of a broader set of challenges to safety; 

• Provide a means for prioritizing these challenges based on risk significance, operating 

experience and / or engineering judgment; 

• Facilitate an integrated consideration of a broader set of factors (i.e., defense-in-depth, 

human reliability) to defend against these challenges; 

• Explicitly identify and quantify sources of uncertainty in the analysis; and 

• Provide a means to test the sensitivity of the results to key assumptions. 

Where appropriate, a risk-informed regulatory approach can also be used to reduce 

unnecessary conservatism in purely deterministic approaches, or can be used to identify areas 

with insufficient conservatism in deterministic analyses and provide the basis for additional 

requirements or regulatory actions. 

Prioritizing the various elements of the licensing review of an applicant's submittal, by noting 

areas in the SRP review procedures of higher and lower importance, can also be viewed as an 

identification of the review areas that have more or less value (i.e., effectiveness and 

importance to safety). Therefore, by focusing review resources on areas of the review that are 

the most effective and safety significant, efficiency can also be improved. 
 B-2 

B.2 Scope, Approach and Process Description 

B.2.1 Scope 

The scope of the SRP prioritizing effort includes all SRP chapters. Within each of these 

chapters, only the review procedures were prioritized. The regulatory requirements and their 

acceptance criteria contained in each chapter were not prioritized, since these need to be met 

regardless of the priority of its corresponding review procedure. 

B.2.2 Approach 

The approach used in developing the prioritization process is a graded approach that combines 

likelihood or consequence insights with deterministic considerations and operating experience. 

It is directed to assess the relative value of performing each review procedure and results in a 

qualitative prioritization considering: 

1) The likelihood of the applicant's non-compliance with a review procedure in the SRP. 

2) The perceived “value added” provided by the NRC review of a given SRP procedural 

step. 

3) The potential consequence if the non-compliance were to remain undetected and 

uncorrected. 

4) The impact on defense-in-depth if the non-compliance remains undetected, assuming 

the review procedure being prioritized was related to a defense-in-depth item. 

Likelihood or consequence insights are those associated with the perceived 

radiological exposure to workers as well as to the public. 

 

The prioritization was done on a generic basis (i.e., no specific dry cask design being 

considered) using the SRP review procedures identified for prioritization. However, it is always 

possible that a design being reviewed will have such unique features (e.g., new material, new 

configurations) that the prioritization needs to be revisited. This can be done on a case-by-case 

basis by reapplying this process on an actual application. 

Finally, in developing the prioritization approach and process, certain assumptions were 

developed . These assumptions included: 

• The cost of correcting a noncompliance was not a factor included in the process. 

• The time and resources required to perform a review procedure were not factors 

included in the process. 

• Dose thresholds used in this process were consistent with thresholds established in 

10 CFR 20 and 10 CFR 72.104. 

• The ”value added” by the review was consistent with the current review level of effort 

and staff experience. 

• Items to be prioritized were chosen such that overlap between them is minimized. 
 B-3 

• All other requirements, except those included in the specific SRP review procedure 

being prioritized, were assumed to be satisfied. 

B.2.3 Process Description 

The process was applied to each technical discipline area in the SRP. The process was 

implemented by the NRC staff reviewers responsible for that discipline (i.e., multiple reviewers 

participated in the prioritization of each review procedure, and the final priority was developed 

based upon a consensus among the reviewers). The process involved looking at each SRP 

review procedure paragraph (or group of paragraphs) in each technical discipline area, and 

asking a structured set of questions. These questions addressed: 

• What is the likelihood of the applicant not meeting the requirement(s) contained in the 

SRP review procedure being prioritized (need for staff review)? 

• What is perceived value added by the staff review (i.e. likelihood of identifying a noncompliance for a given review procedure). 

• What is the potential consequence to public and/or worker radiological safety if the 

requirement(s) remain unmet? 

• What is the impact on defense-in-depth, if any, if the review procedure remains unmet? 

The answers to the above questions were based upon the judgment of the NRC staff reviewers 

who participated in the prioritization process. This judgment reflected the reviewer’s experience 

with current and previous applications and their views regarding potential future problems. 

NUREG-1864, “A Pilot Probabilistic Risk Assessment of a Dry Cask Storage System at a 

Nuclear Power Plant” was previously developed to assess the risk to the public of a specific dry 

storage system at a boiling water reactor site to postulated events. The PRA information was 

not explicitly used in this SRP prioritization because it was limited in scope and assumed that 

the cask was properly designed, constructed and tested. Furthermore, the PRA did not address 

the factors listed in Table B-1 and B-2. It only assessed the risk during cask use from external 

hazards (e.g., fire) and operational errors (e.g., cask drop). Some of these accident sequences 

were also outside the scope of regulatory accidents typically evaluated under Part 72 for 

certified cask systems. In summary, the prioritized review procedures in the SRP address cask 

design, construction and testing, operations, and performance under normal and accident 

conditions to verify compliance with 10 CFR Part 72. 

The steps the reviewers took in prioritizing each SRP review procedure were the following. 

First, the answers to the first two questions were qualitatively determined using a 5 tier 

qualitative ranking. Second, the answer to the third question was qualitatively determined using 

a 3-tier qualitative ranking system. The ranking systems are defined in Tables B-1, B-2 and B3. The quantitative values used in Tables B-1, B-2 and B-3 are intended to serve as guidance 

in the selection of the appropriate qualitative ranking and reflect conservative estimates so as to 

provide a margin to account for uncertainties. The qualitative rankings resulting from Tables B1, B-2 and B-3 were then assigned point values as shown in Table B-4. The point values 

corresponding to the qualitative rankings from Tables B-1, B-2 and B-3 were added together 

and, using the guidance described in Table B-4, an overall qualitative risk component of the 
 B-4 

prioritization (High, Medium or Low) was determined. The reason the scores from Tables B-1, 

B-2 and B-3 were added is that each is a reflection of the importance of the NRC staff 

performing the review procedure being prioritized. Finally, the answer to the last question 

(defense-in-depth) was qualitatively determined using a 3-tier scale (High, Medium or Low) 

following the guidance contained in Table B-5 and Attachment 2 and the reviewer's expert 

opinion. 

At the end of this prioritization process the scoring of all Question 2 were revisited. If the score 

was low or very low (likelihood of staff finding applicants error) it was compared to the LOC 

score (Question 3). If the LOC score was medium or high, then the staff would reevaluate the 

appropriateness of original scoring for Question 2 and would consider raising the score to be at 

least commensurate with the LOC score (Question 3). If the LOC score was low, then the 

significance of the staff failing to discover the mistake is of little consequence. 

The result was a prioritization based on likelihood or consequence and, if applicable, a defensein-depth prioritization ranking. The final prioritization for the SRP review procedure was the 

overall LOC ranking and, if also related to defense-in-depth, a weighed combination of these 

two, with the weights determined by the NRC staff. These weights were determined for each 

review procedure prioritized and used only for that respective item (i.e., the importance of LOC 

versus defense-in-depth may vary from item to item). Attachment 1 to this appendix lists the 

detailed steps associated with implementing the prioritization process that was used in 

assessing the priority of each SRP review procedure. Attachment 2 provides a more detailed 

discussion on defense-in-depth. Attachment 3 provides an example of the documentation and 

major considerations associated with implementation of the process for one specific review 

procedure. 

B.3 SRP Priority Designation and Implications 

Upon completion of the prioritization process, the priority (HIGH, MEDIUM or LOW) associated 

with each review procedure has been indicated in the SRP at the beginning of each paragraph 

in the review procedures. 

The prioritized procedures are intended to ensure that reviews are adequately focused on areas 

that have the most significant impact on safety and compliance with regulatory limits. It is 

important to remember that the priority designations were developed on a generic basis and 

may need to be adjusted depending upon the characteristics of specific applications. It is the 

responsibility of the individual reviewer to assess the design and determine the ultimate rigor 

needed to make a safety determination, with reasonable assurance, in each review area. 

Finally it should be noted that a low or medium priority review procedure does not mean an 

application is exempted from any associated regulatory requirement, design requirement, or 

safety analyses that is expected within the review objectives and acceptance criteria. 
 B-5 

Table B-1 Likelihood of Applicant's Non-Compliance with the SRP Review Procedure 

Likelihood of 

Not Meeting the Requirements Description 

Very High Qualitative: Likely to occur. 

Quantitative: P > 0.5 

High Qualitative: Probably will occur. 

Quantitative: 0.1 < P < 0.5

Medium Qualitative: May occur. 

Quantitative: 0.03 < P < 0.1

Low Qualitative: Unlikely to occur. 

Quantitative: 0.01 < P < 0.03 

Very Low Qualitative: Occurrence improbable. 

Quantitative: P < 0.01

P = Probability 

Table B-2 Potential “Value Added” Through the NRC Review Process 

Likelihood that the NRC Review of a 

Specific Review Procedure Step Will 

Identify a Non-Compliance 

Description 

Very High Qualitative: Likely to occur. 

Quantitative: P > 0.5 

High Qualitative: Probably will occur. 

Quantitative: 0.1 < P < 0.5

Medium Qualitative: May occur. 

Quantitative: 0.03 < P < 0.1

Low Qualitative: Unlikely to occur. 

Quantitative: 0.01 < P < 0.03 

Very Low Qualitative: Not probable. 

Quantitative: P< 0.01

P = Probability 
 B-6 

Table B-3 Potential Impact if the Noncompliance Were to Remain Uncorrected 

Increase in Likelihood or 

Consequence (LOC) 

 if Requirements Remain Unmet 

Description 

High Qualitative: Likely to occur or significant 

consequences.

Quantitative: >10-3/yr* or >25 rem to worker or 

> 1 rem to public. 

Medium Qualitative: May occur or moderate 

consequences.

Quantitative: <10-3/yr but >10-5/yr** or 5 -25 rem 

to worker or 0.1 rem - 1 rem to public. 

Low Qualitative: Occurrence improbable or minimal 

consequences. 

Quantitative: < 10-5/yr or less than 10 CFR 20 

dose limits for workers and the public. 

* 10-3/yr corresponds to the likelihood of an event that could occur in one or more casks 

over a 20 year life of 50 casks. 

** 10-5/yr corresponds to the likelihood of an event that could occur in one or more casks 

over a 20 year life of 5000 casks (i.e., 50 at each of 100 operating reactors). 

Table B-4 Overall LOC Ranking 

Numerical values for Tables B-1, B-2 and B-3 are assigned as follows (note that Table B-3 only 

assigns values of 1 through 3): 

Very High 4 

High 3 

Medium 2 

Low 1 

Very Low 0 

For each SRP review procedure, the qualitative scores from Tables B-1, B-2 and B-3 are added 

and a combined qualitative score is determined as follows: 

High 9 - 11 

Medium 6 -8 

Low 1 - 5 
 B-7 

Table B-5 Defense-in-Depth Ranking 

Defense-in-depth has long been a key element of the NRC’s safety philosophy. It is intended to 

ensure that the accomplishment of key safety functions is not dependent upon a single element 

of design, construction, maintenance or operation. In effect, defense-in-depth is used to provide 

one or more additional measures to back up the front line safety measures, to provide additional 

assurance that key safety functions will be accomplished. Traditional defense-in-depth 

measures for reactors have included items such as confinement, containment, redundant and 

diverse means of decay heat removal and emergency evacuation plans. For DSS, examples of 

measures associated with defense-in-depth are discussed in Attachment B-2. Defense-in-depth 

measures are generally decided upon using deterministic considerations (i.e., engineering 

judgment) regarding the importance of the safety function and the potential uncertainties that 

could affect its performance. 

With respect to prioritizing the review procedures in this SRP, a review procedure can be 

considered associated with defense-in-depth if it is related to providing a backup to the front line 

of defense (e.g., confinement is generally considered a defense-in-depth measure since it 

provides a backup to cladding integrity). 

Defense-in-depth measures are not intended to detract from the importance of front line safety 

measures. Defense-in-depth measures are intended to provide additional assurance so the 

safety function can be accomplished. It is not the intent of defense-in-depth to reduce the 

importance of the front line safety measures since, if their importance were reduced, the 

importance of the NRC staff review associated with those measures could also be reduced, 

which could affect the reliability or performance of the front line safety measures. This could 

leave the defense-in-depth measures as the primary means of performing the safety functions, 

instead of being the backup. 

If failure to perform the review procedure could impact defense-in-depth (assuming the front line 

safety measure has failed) and has: 

• a low likelihood and/or consequence, then the paragraph should be prioritized as "LOW." 

• a medium likelihood and/or consequence, then the paragraph should be prioritized as 

"MEDIUM." 

• a high likelihood and/or consequence, then the paragraph should be prioritized "HIGH." 

Likelihood and consequence are defined in Table B-3. 
 B-8 

Attachment B-1 

Process Steps to Prioritize SRP Review Procedures 

The following steps should be followed in prioritizing each review procedure. Multiple staff 

reviewers in each technical area should participate in the prioritization so as to arrive at a 

consensus on the priority. The checklist at the end of this attachment can be used to document 

each step. 

1. Identify the SRP review procedures to be prioritized, with a focus on the requirements 

that the review procedure is checking. This will result in individual paragraphs (or 

groups of paragraphs) being prioritized as separate items. 

2. Estimate the likelihood that the requirement related to the SRP review procedure will not 

be met by the applicant by choosing the appropriate likelihood range from Table B-1 

(Likelihood of Applicant's Non-Compliance with the SRP Review Procedure). This 

estimate can be affected by several factors, including the experience of the applicant, 

the novelty of the technology used in the application, the difficulty level of meeting the 

requirement, the applicant's quality assurance program, etc. 

 The rankings listed in Table B-1 are arranged to provide more staff review effort where it 

is determined that the applicant is less likely to meet the review procedure. Conversely, 

where it is felt that the applicant will meet the review procedure, less staff effort would be 

required. 

3. Estimate the likelihood that if the requirement is not met, this fact will be discovered by 

performing the SRP review procedure. This is done by choosing the appropriate 

likelihood range from Table B-2 (Likelihood that the NRC Review Would Identify the 

Non-Compliance, Given that it Exists). This factor may be relatively high, however, there 

may be review procedures that have varying degrees of implementation. 

 The rankings listed in Table B-2 are arranged to continue to provide a high level of staff 

effort in areas where the staff review has typically identified problems. Conversely, 

where historical staff review efforts have not identified problems, that level of staff effort 

is minimized. 

4. Estimate the potential radiological likelihood or consequence to public and worker safety 

if the requirement were to remain unmet. It is recognized that this is not a trivial task and 

that no complete probabilistic risk assessment (PRA) is available for dry casks or 

ISFSIs. The following was intended to aid the prioritizer with this assessment: 

• Consider potential event sequences or a set of event sequences, such that the 

dose to the most exposed person from these sequences includes the bulk of the 

dose from all possible sequences. The premise here is that every possible 

sequence of events has some likelihood of occurring and results in some dose to 

workers and the public. Some sequences are very likely and result in very little 

dose, others are very unlikely and result in very large dose, etc. The prioritizer 

should use experience in considering the sequence(s) that have the highest 

radiological likelihood or consequence to the most exposed person. This is 

equivalent to answering the following questions: 
 B-9 

- What can happen? (i.e., what can go wrong?) 

- How likely is it that that will happen? 

- If it does happen, what are the consequences? 

• Using Table B-3 Potential Impact if a Noncompliance is not identified, determine 

the corresponding range of increased likelihood or dose. This range corresponds 

with the likelihoods or consequences for the dominant sequences. 

The rankings listed in Table B-3 are weighted to devote more staff resources to the 

review procedures that are viewed to be more likely or consequence significant and less 

staff resources to those that are viewed to be less likely or consequence significant. 

5. The prioritizer now has three qualitative rankings corresponding to: 

• Likelihood of the applicant not meeting the requirements. 

• Likelihood that the NRC Review would find the discrepancy, given that it exists. 

• Potential consequences if the requirements remain unmet. 

Using these three rankings, determine the overall qualitative LOC (Likelihood or 

Consequence) ranking (High, Medium or Low) for this review procedure by adding the 

numerical values assigned to each qualitative ranking and the guidance in Table B-4. 

6. Using Table B-5, assess the applicability and impact on defense-in-depth, if any, if the 

SRP review procedure is not met. Defense-in-depth consists of a number of elements 

as discussed in Attachment 2 and will not be applicable to all review procedures. If 

applicable, this step results in a High / Medium / Low qualitative ranking. 

 

7. There is now a qualitative ranking and, if applicable, a qualitative defense-in-depth 

ranking. The method of combining these scores reflects the relative importance given to 

risk versus defense-in-depth. Judgment must be used to integrate these two rankings 

into a single ranking applicable to the SRP review procedure. This integration is done by 

weighing the two rankings using weights determined by the NRC reviewers. The 

weights are determined for each review procedure being prioritized and used for that 

procedure only. 

8. A prioritization process checklist is to be filled out for each paragraph (or group of 

paragraphs) prioritized, so as to document the basis for the priorities assigned to each 

review procedure. This checklist is shown on the following page and Attachment B-3 

provides an example of a completed checklist for a specific review procedure. 
 B-10 

 Prioritization Process Checklist

 Chapter: Paragraph Number: 

STEP SCORE COMMENTS 

 

1. Identify the SRP procedure 

to be prioritized. 

N/A 

 

2. Likelihood that requirement 

will not be met (Table B-1). 

 

 

3. Likelihood that staff reviews 

will find discrepancy 

(Table B-2). 

 

 

4. LOC if requirement is not 

met (Table B-3). 

 

 

5. Determine combined LOC 

value (Table B-4). 

 

 

6. Determine defense-in-depth 

value (Table B-5), if applicable. 

 

 

7. Determine relative weight of 

risk and defense-in-depth 

values determined in (steps 5 

and 6 above). 

 

 

8. Overall priority (Combine 

LOC and defense-in-depth 

values). 

 

 
 B-11 

Attachment B-2 

Defense-in-Depth (DID) 

Defense-in-depth has long been a key element of NRC’s safety philosophy. It is intended to 

ensure that the accomplishment of key safety functions is not dependent upon a single element 

of design, construction, maintenance or operation. In effect, defense-in-depth is used to 

compensate for uncertainties by employing one or more additional measures to back up the 

front line safety measures, thus providing additional assurance that key safety functions will be 

performed. Traditional defense-in-depth measures for reactors have included items such as 

confinement, containment, redundant and diverse means of decay heat removal and emergency 

evacuation plans. Defense-in-depth measures are generally decided upon using deterministic 

considerations (i.e., engineering judgment) regarding the importance of the safety function and 

the potential uncertainties that could affect its performance. 

In the dry cask SRP prioritization, each paragraph (or group of paragraphs) to be prioritized, 

would be examined individually from a DID perspective to determine if that paragraph (or group 

of paragraphs) is related to defense-in-depth. If so, and if the paragraph is not met, a 

determination would then be made as to whether or not a defense-in-depth measure could be 

compromised and the risk significance. 

To determine if a defense-in-depth measure could be compromised, it is first necessary to 

decide what are defense-in-depth measures? To help make this decision, the following 

guidance was used. 

• A defense-in-depth measure is any design feature or action that is required by the SRP 

as a backup measure to the front line safety measures. This ensures that, if the front 

line safety measure is lost, the backup measure is present to perform that safety 

function. 

DSS defense-in-depth measures may include: 

• Confinement System (2nd barrier to fuel clad integrity); 

• Operating Controls and Monitoring 

• Non-mechanistic and bounding event analyses (to mitigate site-specific uncertainties). 

SRP review procedures that relate to items that can be considered defense-in-depth should 

receive a DID ranking. 

If the SRP paragraph (or group of paragraphs) being prioritized is related to a measure that 

meets the above guidance, then it would be evaluated as a defense-in-depth measure and 

prioritized as follows: 

• If the failure of the front line and DID measures relative to the issue identified in the SRP 

review procedure would result in a low likelihood and / or consequence, then the 

paragraph should be prioritized as "LOW." 
 B-12 

• If the failure of the front line and DID measures relative to the issue identified in the SRP 

review procedure would result in a medium likelihood and / or consequence, then the 

paragraph should be prioritized as "MEDIUM." 

• If the failure of the front line and DID measures relative to the issue identified in the SRP 

review procedure would result in a high likelihood and / or consequence, then the 

paragraph should be prioritized "HIGH." 

Risk and consequence are defined in Table B-3. 

It should be noted that defense-in-depth measures are not intended to detract from the 

importance of front line safety measures. Defense-in-depth measures are intended to provide 

additional assurance so the safety function can be accomplished. 
 B-13 

Attachment B-3 

This attachment provides an example of a completed prioritization checklist to illustrate the level 

of documentation and major considerations associated with the prioritization of each specific 

review procedure. The review procedure used in the example is Section 4.5.4.7 "Confirmatory 

Analysis" in Chapter 4 "Thermal Evaluation" of NUREG-1536. A total of three staff reviewers 

participated in the prioritization of Chapter 4 and the prioritization input and outcome reflects a 

consensus among the reviewers. 

Prioritization Process Checklist

 Chapter: 4 - "Thermal Evaluation" Paragraph Number: 4.5.4.7 

STEP SCORE COMMENTS 

 

1. Identify the SRP procedure 

to be prioritized. 

N/A Done by reviewers. 

 

2. Likelihood that requirement 

will not be met (Table B-1). 

L Applicant provides calculations using 

generally accepted analytical tools. 

3. Likelihood that staff reviews 

will find discrepancy 

(Table B-2). 

H Staff provides a thorough review. 

 

4. Risk if requirement is not 

met (Table B-3). 

H Fuel cladding (i.e., first line-of-defense 

for fission product retention) could fail if 

thermal analysis is incorrect. 

5. Determine combined risk 

value (Table B-4). 

M L (1) + H (3) + H (3) = 7 (MEDIUM) 

6. Determine defense-in-depth 

value (Table B-5), if applicable. 

H Provides independent check (i.e., 

second line-of-defense) as backup to 

front line staff review of applicant's 

submittal. 

 

7. Determine relative weight of 

risk and defense-in-depth 

values determined in (steps 5 

and 6 above). 

DID 

> Risk

DID is more important than risk since it 

has the potential to uncover applicant or 

staff review errors and can provide 

additional insights for probing the validity 

of the applicant's analysis. 

 
 B-14 

STEP SCORE COMMENTS 

8. Overall priority (Combine 

risk and defense-in-depth 

values). 

H DID controls final priority. 

 
C-1

APPENDIX C 

INTERIM STAFF GUIDANCE (ISG) INCORPORATED INTO NUREG-1536 Revision 

ISG # 

& Rev. Title NUREG 1536 Revision 1 Status

ISG 1 

Rev. 2 

Damaged Fuel Added 

ISG 2 

Rev. 1 

Fuel Retrievability Added 

ISG 3 Post Accident Recovery and Compliance with 

10 CFR 72.122(l) 

Added 

ISG 4 

Rev. 1 

Cask Closure Weld Inspections Superseded by ISGs 

15 and 18 

ISG 5 

Rev. 1 

Confinement Evaluation Added 

ISG 6 Establishing Minimum Initial Enrichment for the Bounding 

Design Basis Fuel Assembly(s) 

Added 

ISG 7 Potential Generic Issue Concerning Cask Heat Transfer in a 

Transportation Accident 

Added 

ISG 8 

Rev. 2 

Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks 

Not Added, but 

referenced 

ISG 9 

Rev. 1 

Storage of Components Associated with Fuel Assemblies Added 

ISG 10 

Rev. 1 

Alternatives to the ASME Code Added 

ISG 11 

Rev. 3 

Cladding Considerations for the Transportation and Storage of Spent Fuel 

Added 

ISG 12 

Rev. 1 

Buckling of Irradiated Fuel Under Bottom End Drop Conditions 

Added 

ISG 13 Real Individual Added ISG 14 Supplemental Shielding Added 
C-2

ISG # 

& Rev. Title NUREG 1536 Revision 1 Status

ISG 15 Materials Evaluation Added ISG 16 Emergency Planning NA ISG 17 Interim Storage of Greater Than Class C Waste NA 

ISG 18 

Rev. 1 

The Design & Testing of Lid Welds on Austenitic Stainless Steel Canisters as Confinement Boundary for Spent Fuel Storage 

Added 

ISG 19 Moderator Exclusion Under Hypothetical Accident 

Conditions and Demonstrating Subcriticality of Spent Fuel Under the Requirements of 10 CFR 71.55(e) 

NA 

ISG 20 Transportation Package Design Changes Authorized Under 

10 CFR Part 71 Without Prior NRC Approval 

NA 

ISG 21 Use of Computational Modeling Software Added 

ISG 22 Potential Rod Splitting Due to Exposure to an Oxidizing 

Atmosphere During Short-Term Cask Loading Operations in LWR or Other Uranium Oxide Based Fuel 

Added 

ISG 23 

(Draft) 

Draft - Application of ASTM Standard Practice C1671-07 when performing technical reviews of spent fuel storage and transportation packaging licensing actions 

Not Added 

ISG 24 

(Draft) Reserved N/A 

ISG 25 

(Draft) 

Draft - Pressure and Helium Leakage Testing of the Confinement Boundary of Spent Fuel Storage Casks 

Added 

ISG 26 

(Draft) 

Reserved N/A 
D-1

APPENDIX D PUBLIC COMMENTS RECEIVED AND THEIR DISPOSITION 

The purpose of this appendix is to list all the public comments received on NUREG-1536 “Standard Review Storage Systems at a General License Facility,” Revision 1A. The NRC issued NUREG-1536, Revision 1A public comment on April 15, 2009 for a 90 day period and received comments from the following two sources: 

• NEI, Nuclear Energy Institute, letter to Mr. Ron Parkhill, USNRC, dated July 14, 2009 (ML 091970430) • NAC International, email from Mr. Tony Patko to Mr. Ron Parkhill, USNRC, dated July 15, 2009 (ML 092

The staff’s resolution and any associated changes to the standard review plan are listed for each commenumbers listed in the attached table refer to the line numbering of Revision 1A of NUREG-1536. 
D-2

NUREG-1536 Public Comments and Resolution 

Comment SRP Location Public Comment Resolution Cha

NEI 1 General The SRP discusses the 

content of the Technical Specifications in numerous locations. While the NRC does not have a policy statement on technical specifications for dry cask storage systems, the NRC Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, as published in the Federal Register at 58 FR 39132, July 22, 1993, provides useful guidance. The Final Policy Statement discusses, in the Background, the trend towards adding information to the Technical Specifications by stating: 

“... since 1969 there has been a trend towards including in technical specifications not only those requirements derived from the analyses and evaluation included in the plant's safety analysis report but also essentially all other NRC requirements governing the operation of nuclear power plants. ... In the Commission's view, this has diverted both NRC 

The staff recognizes the policy statement for operating reactors. The staff also notes that sitespecific operating reactors are different than certified dry cask systems. Reactors represent an inherently higher risk from accidents, and maintain several active systems and active monitoring of key performance parameters during operations, Reactor technical specifications (TS) are established to ensure these functions are maintained in order to ensure adequate containment, reactivity control and thermal hydraulic control of the system during operations. Dry storage casks are passive in nature, and do not typically rely upon multiple active safety systems to mitigate events during storage operations. Instead they rely on passive design features and administrative controls to assure criticality safety, confinement safety, and cladding protection during normal, off-normal, and accident conditions. The staff considers TS to be valuable, in part, to assure the most important fabrication, design features, 

Chapter 13 If a reviewedesign featspecificatioassumptionassumptionoperation, programmaSAR item is should not NRC staff ashould be fconsidered condition ospecificatioshould coninsights, saoperational in-depth conovelty, anare unique design. Thalso implemthis chapteconditions specificatio
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Comment SRP Location Public Comment Resolution Cha

staff and licensee attention from the more important requirements in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety.” 

The Final Policy Statement also stated: 

“The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval.” 

A similar philosophy where only those items that have a direct nexus to the protection of the public health and safety from an immediate threat are included in the Technical specifications should be 

contents, and operations of the system are appropriately controlled among the diversity of site users, to mitigate the likelihood and consequences of potential offnormal and accident conditions. The dry cask storage certificate includes a condition which specifies TS. These control the fabrication, safe use, and operation of the dry cask system during loading, transfer, and passive storage. This is consistent with the Commission’s policy statement published in Federal Register, 58 FR 39132, July 22, 1993. The staff also believes the format typically employed for DCSS TS is amenable to the use by general licensees into assuring safe use and operation. Several factors may influence the content of TS. Chapter 13 is revised to clarify these factors. The staff finally notes that NEI has identified this as a future issue to be discussed between the NEI dry storage task force and NRC requested to discuss concerns with cask TS in a separate interaction with NRC (see ML093310122 ). If a new philosophy were adopted, 
D-4

Comment SRP Location Public Comment Resolution Cha

adopted. The guidance to the staff in the draft SRP in regards to Technical Specifications should be revised accordingly. 

or generic changes were made to dry cask technical specifications, then the standard review plan may be subject to future revision to implement any associated guidance with the changes. 

NEI 2 General The document should state 

throughout that for canisterbased systems the “confinement cask” is the welded canister assembly. 

The SRP is written to apply to both bolted casks as well as to welded canister systems. The term confinement cask applies to both the bolted cask, as well as, welded canister. SRP Section 5.5.1.2. specifies the design and qualification guidance for a welded canister to qualify as a confinement boundary 

No Change 

NEI 3 General In a number of locations, the 

guidance gets into specifying the details of the ASME Code and other codes. Unless NRC does not accept what the codes require, the guidance should avoid repeating the code details and simply refer to the code at a higher level (e.g., “Section III, Subsection NB”). 

Specific guidance is provided in certain instances to avoid misunderstanding and possible conflicting interpretations. The specific guidance also assists reviewers in focusing on important elements of the ASME Code with respect to the associated review objectives. 

No Change

NEI 4 General Renumbering the chapters in 

the SRP may create confusion during future licensing actions where the SRP chapters will not coincide with the SAR chapters. Please consider restoring the current SRP revision chapter numbering 

This revision of the SRP included the addition of a new Materials chapter and the deletion of the Decommissioning chapter which affected the numbering of many of the chapters. New certificate requests may follow the new format, and amendments to an 

No Change 
D-5

Comment SRP Location Public Comment Resolution Cha

sequence. existing certificate may follow the 

format as licensed. However, NRC intends to revise associated Regulatory Guides that specify acceptable content and format of certificate applications. The revised draft regulatory guides will also be issued for public comment. 

NEI 5 34 The statement that ISGs were 

developed to address changes in requirements differs from the definition of ISGs provided at line 660. This statement should be consistent with line 660 to avoid implying that ISGs impose new requirements as could be interpreted by the current wording. 

The staff agrees the wording should be more consistent and the SRP is revised as appropriate. The statements were not intended to imply that ISGs impose new regulatory requirements, because the SRP is only for guidance to staff. 

Changed a“These ISGclarify imporegulatory rlessons leatechnologydetailed tecAlso changsentence oaccordingly

NEI 7 542 Editorial: Change “term” to 

“terms.” 

Agree with comment Changed to 

NEI 8 542 

791 

1259 4522 4993 6853 

Change “containment” to “confinement” to use more storage-specific language. 

Agree with comment Changed th

confinemencontainmen

NEI 9 541-542 

8319-8320 

It is not clear why peak rod average burnup is included in this definition and later in the SRP. Assembly average burnup is typically used for specifying allowable contents and should be sufficient 

Agree with comment that assembly average is typically used for specifying allowable contents. In addition, the peak rod average burnup is a parameter considered in the fuel integrity analyses in the materials review. 

Definition cthat assemis used for content, anaverage buassessing fin the mateexception a
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Comment SRP Location Public Comment Resolution Cha

8.4.17. 

NEI 10 605-606 Revise definition to account for 

a DFC that could contain less than one assembly (e.g. failed rod basket with 50 rods vs. 264 for an assembly) or more than one assembly for a consolidated rod can. Suggest “A metal enclosure to confine damaged spent fuel. A damaged fuel can with its damaged spent fuel contents must satisfy ...” 

This is the definition currently in ISG-1 Rev 2. Because the ISG also applies to the transportation SRPs, the definition will be maintained for consistency at this time. However, this does not preclude an applicant proposing the use and evaluation of a DFC that may contain fuel rods that are more or less than that associated with one fuel assembly. 

No change 

NEI 11 667-669 a) M.O.S. is not “identical” to 

F.O.S. b) “M.O.S” in the first set of parentheses should be “F.O.S.” c) Line 669: delete the first occurrence “-1” , 

Agree with comment. Margin of safety was restated in the document as suggested. 

Text changmay be deffactor of sacapacity/deF.S. (capac(with minimMofS > 0.0 

NEI 12 684 In the 2nd sentence, add 

“neutron” between “high” and “absorption.” 

Agree with comment. Revised to 

absorption” 

NEI 13 687 Suggest deleting “and 

transporting” because this SRP is exclusively for storage. 

Agree with comment. Definition c

reference t“storage”. 

NEI 14 720 A definition is provided for 

BPRA at line 532 but definitions are not provided for control element assemblies (CEAs) or thimble plug assemblies (TPAs). 

Agree with Comment The definiti

revised to idefinitions f

NEI 15 740 While preferential loading is 

currently used for thermal 

Agree with comment. Definition c

A non-unifo
D-7

Comment SRP Location Public Comment Resolution Cha

loading, it is also used for dose reduction and could be used in the future for other reasons (e.g., criticality control). This definition should be more flexible. 

configuratioassemblies system thaspecified by zone desigbasket cell, limiting nucparameters that can be zone. Prefoften used optimize alparameters cooling timsatisfying thcriticality, aperformanccask system

NEI 16 748-752 The definition of “Ready 

Retrievability” is incorrect and inconsistent with Section 12.4.5 (lines 11208 – 11219) of the SRP and draft ISG-2 Rev 1 which has been issued by the NRC for comment. The first sentence of this definition is the definition of recovery not retrievability. This definition should be revised and a definition for “recovery” should be added. 

Agree with comment. ISG-2, Rev 1 was issued as final on February 22, 2010 (ML100550817). The ISG considered and addressed public stakeholder comments. This SRP has been administratively updated to incorporate Rev 1 of ISG-2 Definitions for “ready retrieval” and “normal means” have been added in accordance with the ISG-2 Rev. 1 guidance. The definition for “retrievability” is changed to be consistent with the language of 10 CFR 72.122(l). A definition for “recovery” has been added. The SRP is also revised to use 

Changed dRetrievabiliNormal Me“Retrievabi

with 10 CFsystems mallow readfuel, higwaste, aGTCC wprocessing Ready retri

move a canspent fuel t
D-8

Comment SRP Location Public Comment Resolution Cha

consistent terminology for retrievability and ready retrieval in sections 1, 2, 8, and 12. The term “ready retrievability” has been change to “retrievability” or “ready retrieval” as appropriate.. 

transportatlocation whcan be remretrieval alsmaintaining individual oassemblies normal meaNormal memove a fuecontents by and grapple undamaged point of casaddition of modificatioto make the for lifting by does not pras being coby normal mRecovery 

return the material to after an acendangerinsafety. Thensuring release of to the enviexposures the limits during pos
D-9

Comment SRP Location Public Comment Resolution Cha

operations.

NEI 17 810 Clarify this definition to say 

that the supplemental shielding is only ITS if it is credited in the 72.104 dose analysis. 

Agree with comment. Changed to

Supplemendeemed as important to specified in Specificatiouse of the sif credited iradiation prfor meeting 72.106(b) r

NEI 18 892-895 

2358-2359 6623-6625 7261-7266 

7286 

7318-7319 7413-7414 7456-7457 

11350-11351 

11366 

11454-11461 11513-11516 11521-11523 11527-11597 

The bases for what requirements should be in the CoC or TS provided in these sections are vague, subjective, not risk-informed, and not consistent with practice in NRR (i.e., Part 50 TS). Examples: a) “Any aspect of the design or procedures that the NRC determines should not be changed” (892-895) b) “preclude the possibility of damage to the structure or damage to the confined nuclear material” (2358-2359) c) “any technical aspect of the design which is deemed critical to nuclear safety” (7318-7319) d) whatever “the staff deems necessary” (11350 – 11351) e) “a reviewer deems an item 

See resolution to NEI1 Technical specifications are part of the CoC, and specific guidance remains important for limiting parameters or procedures in these areas to ensure safety of the system during normal operations and accident conditions. There is a diversity of dry cask storage technologies, which employ different types of design features and analytical methodologies to ensure safety with different safety margins calculated within each discipline. Cask technologies continue to evolve with innovative, first-of-akind approaches to ensure confinement, shielding, and criticality safety. In addition, vendors have proposed a diversity 

See changNEI1. In addition, made as a comment intechnical spidentified vSection 8.2 Comment v8.4.1. Alsotechnical spSection 8.4Procedures Criteria, of version was it was not asection, noincorrect. 
D-10

Comment SRP Location Public Comment Resolution Cha

so important” (11366) Given that these casks are loaded and operated at NRClicensed Part 50 facilities, we suggest SFST adopt a function-based, risk-informed set of criteria for what information belongs in the CoC and TS, similar to 10 CFR 50.36(c) for power reactors, recognizing the passive design and operation of storage casks and modules. In general, the TS should only cover operational items under the user’s control for implementation, and only critical design features under the control of the CoC holder, similar to those in the “Design Features” section of Part 50 TS. Examples of information not appropriate for inclusion in TS: fuel basket dimensions (line 6624); alternate materials and other material requirements (7261-7266, 7456-7457); QA/QC documents, procedures, and test protocols for neutron absorbers (7413-7414); ASME Code information (11454-11461), and training (11521-11523). Including this information only 

of TS in terms of scope and format to both assure that safety is maintained and to satisfy specific operational needs of general licensees. All of these factors have contributed to the diversity of TS formats; as well as some of the generality specified for TS in the SRP guidance. The staff recognizes that if applied correctly 72.48 may be used to evaluate if NRC approval is needed for changes. However, 72.48(c)(1)(B) itself, recognizes the role of certificate conditions and TS in limiting design changes without NRC approval. These certificate conditions and TS areas established at the discretion of NRC. The staff finally notes that NEI has identified this as a future issue to be discussed between the NEI dry storage task force and NRC (see ML093310122). If a new philosophy were adopted or generic changes were made to dry cask technical specifications, then the standard review plan may be subject to future revision to implement any associated 
D-11

Comment SRP Location Public Comment Resolution Cha

in the FSAR is appropriate based on risk. 72.48 provides adequate controls for determining whether prior NRC approval is required for changes to these items, and the QA program adequately addresses training and manufacturing. It is also a poor practice from a human factors standpoint to incorporate portions of the FSAR into the CoC by reference. 

guidance with any change. Further resolution of this issue through the SRP comment resolution process would not be practical at this time. 

NEI 19 1255 Suggest the word “removed” 

instead of “retrieved”. The damaged fuel container is used to assist in placing and removing damaged fuel from the canister. 

Agree with comment. The last se

1.5.1 was m“Therefore, verify that tcontains a the damagecanned, the the can, anwhich the cinserted intthe cask.” 

NEI 20 1259 

11524 11527 

Editorial: Add “and Limits” to the title of Chapter 13. 

Agree with comment Changed th

Limits” 

NEI 21 1540-1541 The operational history 

parameters need to be reasonable values assumed in the depletion calculations and not bounding values the user must verify that their reactor history meets 

When using burnup credit, the fuel must be confirmed to meet the bounds of the operational history parameters assumed in the analysis or these parameters must be shown to be sufficiently bounding over the full range of fuel to be authorized for loading. 

No change
D-12

Comment SRP Location Public Comment Resolution Cha

NUREG/CR-6716 provides results 

on a study of the importance of and the sensitivity of K-effective to changes in some major parameters. Section 7.5.5.3 of the SRP provides additional guidance regarding bounding assumptions. 

NEI 22 1552-1553 Delete this bullet. “Inerting 

atmosphere requirements” is not an SNF specification and the maximum number of fuel assemblies is specified two bullets prior. 

Agree with comment. Removed b

NEI 23 1154-1155 Based on the elimination of the 

SAR chapter on decommissioning, consider deleting the sentence regarding the planned decommissioning process. 

Agree with comment. Deleted the 

“Additionallshould be iplanned deprocess.” 

NEI 24 1253 

1600-1601 

2139 2204 

2337-2338 

2401 (flowchart 

box for Chapter 

12) 

2508-2512 

3037 3053 8803 9075 

11314 

These lines are inconsistent with Section 12.4.5. of the draft SRP (lines 11215-11219) and other portions of the SRP which state that retrievability in 10CFR72.122(10CFR72.122 (l) applies only to normal and off-normal conditions and not accident conditions. These lines are also inconsistent ISG-2 Revision 0 and draft ISG-2 Revision 1. Reference to retrievability should be removed in discussions of 

Agree with Comment. See response to NEI 16. Section 12.4.5 is also clarified to discuss the applicability of recovery and retrievability. Retrievability applies to normal and off-normal events, and not design basis accidents. The meaning of normal condition, off-normal events and design-basis accidents, in this context, are further clarified in Section 12.1 

See changThe SRP is consistent tretrievabilitin sections The term “rhas been c“retrievabiliretrieval,” aOther referoff-normal events havdistinguishi
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Comment SRP Location Public Comment Resolution Cha

accident conditions throughout the SRP. 

as “retrievaas applicabSection 12to include “Normal cointended opevents, and conditions, reasonably with high frstorage opeOff-normal man-made phenomena with moderonce per caANSI/ANS events as D

NEI 25 1374 Since the NRC is currently 

working on rulemaking that would change the licensed lifetime of a cask, it is suggested that a reference to the 20 year limit be removed here and throughout the document and that a reference to the regulation be provided instead. 

Agree with Comment Changed S

remove the and referenwhere apprsimilarly ch2.4.3.1, SeSection 8.5 

NEI 26 1704 Identifying the fuel vendor is 

not pertinent to the review and should be deleted. 

It is necessary to distinguish the fuel vendor so that staff can distinguish between the many different types of fuel assembly 

No change
D-14

Comment SRP Location Public Comment Resolution Cha

variations that exist and whose materials properties are not identical. 

 NEI 28 1913-1917 This paragraph is inconsistent 

with ISG-5 (for metal casks) and ISG-18 for welded canisters. Non-mechanistic confinement boundary failures are no longer part of the cask design and licensing basis. 

Agree with comment Removed th

Sentence: “Nassessment demonstrate the storage NRC staff coDSS should effects of a cboundary fa

NEI 29 1992 Editorial: Change “..” to .” Agree with comment Changed aNEI 30 2041 Change “SNF retrieval” to 

“retrievability”. 

Agree with comment Changed a

NEI 31 2110 Change “retrieval capability” to 

“retrievability”. 

Agree with comment Changed a

NEI 32 2271-2280 ANSI/ANS-57.9 is outdated 

and not germane to many of today’s commercial spent fuel systems. Other than the design event classifications, care should be used in referring to this standard for today’s DSS designs. 

The use of ANSI/ANS-57.9 is broader than just event classifications. Each applicant should evaluate and justify the applicability of ANSI/ANS-57.9 to its proposed DSS. The reference to the review standard is reworded to be consistent with current terminology in the structural review chapter. 

Revised the system struISFSI dry sSection 3.4

NEI 33 2278-2280 Editorial: The last sentence of 

this paragraph does not appear to be grammatically correct. 

Agree with comment Changed s

loadings deSociety of C“Minimum DBuildings aStructures,used when 
D-15

Comment SRP Location Public Comment Resolution Cha

are consideANSI/ANS-

NEI 34 2308, 2626 

3085, 3501 

Some inconsistency is noted regarding the specified Code years When referring to the ASME code, no code year was mentioned. However, when referring to a non-ASME code, a code year was mentioned. For example, line 2308, IBC code (2006), line 2626, ASTM C33 (2002), line 3085, ANSI/ANS-57.9 (1992), line 3501, ACI 349 (2006). To avoid confusion and permit appropriate flexibility for the applicant, the code year should not be mentioned in the review plan 

Specific guidance regarding codes year is provided in certain instances to avoid misunderstanding and possible conflicting interpretations. However, this does not preclude an applicant from proposing the use of alternate codes or code years with appropriate justification. 

No Change 

NEI 35 2340, 2713 Regarding Line 2340, “This 

position does not necessarily require that all confinement system and other structures important to safety survive all design-basis accident and extreme natural phenomena without any permanent deformation or other damage” and Line 2713, “The system should not experience any permanent deformation or loss of safety function capability during normal or off-normal operation conditions.  However, the system may 

The NRC staff agrees that the design analysis, in accordance with provisions in Section III of the ASME Code, does not restrict use of linear-material properties. To ensure clarity, the paragraph beginning in Line 3168 was rewritten to recognize the potential inelastic structural behavior for the ASME Code, Section III, Appendix F accident load conditions. 

The entire in the secoSection 3.5changed tothe provisioSection III, inelastic mamay be usecask design for accidenshould idenused for the properties.” 
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Comment SRP Location Public Comment Resolution Cha

experience some permanent deformation, but no loss of safety function capability, in response to an accident” please consider the following: Based on the above discussion, elastic-plastic analysis should be allowed to analyze the accident load; however, Line 3168, “to be consistent with the provision in Section III of the ASME code, the analysis should use linear material properties. For materials that do not serve in structural capacity (such as shielding materials), inelastic material properties may be used for cask components that are not stress-limited and respond inelastically to the load conditions for storage casks” implies that only elastic analysis can be used unless you use strain limited criteria.  In the past NRC has accepted the use of elastic-plastic properties for all the accident load analyses and stress limited criteria are used per ASME Appendix F. 
D-17

Comment SRP Location Public Comment Resolution Cha

NEI 36 2357-2362 The first sentence of this 

paragraph seems to indicate that TSs should be in place to preclude possibility of damage to the structure or the confined material during cask handling and operations. The second sentence of the same paragraph seems to indicate that TS should describe the actions and inspections to be conducted upon occurrence of “events” that may cause such damage. These two statements appear to be contradicting each other. 

In the unlikely event of cask damage resulting from cask handling and/or operation, the second sentence discusses actions and inspections that should be conducted to ensure that the cask is secured in a safe configuration. 

No change 

NEI 37 2380 Editorial: Add a blank line 

between lines 2379 and 2380 

Agree with comment Changed a

NEI 38 2612-2640 This section seems to imply 

that the alternate concrete temperatures described apply only to the steel-lined concrete confinement cask system designed to ACI 359. Similar concrete temperature provisions have been accepted for the NUHOMS HSM type concrete structure designed to ACI 349. 

The text explicitly states that the temperature limits presented as an exception to Section CC-3340 of ACI 359 are temperature limits that apply as an alternative to ACI 349, A.4. The inclusion of steel-lined concrete confinement cask systems is an additional configuration in which alternative temperature requirements can be employed, not the sole configuration for their use. 

No Change 

NEI 39 2621, 8210 Add ASTM C150 as the 

standard specification for Type II cement. 

Agree with comment Changed to 

Satisfy ASTSatisfy AST
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Comment SRP Location Public Comment Resolution Cha

SpecficatioCement”) rother requiin ACI 349 Have demo

NEI 40 2627 Delete “2002” (edition year of 

ASTM C33) 

Agree with comment Rewrite “Sa

(“Standard Concrete Arequiremen

NEI 41 2729 

2735 5908 8889 

Change “retrieval” to “unloading” or “removal” as applicable. 

These lines discuss normal condition operations, so retrieval is the correct word choice for each case, and is essentially synonymous with “unloading” or “removal” in the context it is used. 

No change 

NEI 42 2879-2882 The passage: “The SAR 

should identify the maximum response determined. That response should be sufficiently low such that while damage may occur, it would not impair the capability of the component to perform its safety functions” is not clear. What, specifically, is meant by “maximum response”? 

In the context of structural analysis for the explosive overpressure, the generalized term, “maximum response,” generally means to include pressure induced maximum stresses at critical cask locations and governing structural performance modes for the cask components important to safety. This is added to the SRP to provide clarity. 

Add the folsecond par3.5.1.4 i. (3The maximincludes prmaximum scask locatiostructural pfor the cask important to 

NEI 43 2885 The third paragraph of the 

current SRP version has been deleted in this proposed revision to NUREG 1536. The deleted paragraph accepted the fire parameters from Part 71 as a basis for characterizing the fire during storage. Additionally, it 

Partially agree with comment. The deleted paragraph contains the lead sentence, "The NRC has accepted the fire parameters included in 10 CFR Part 71 as the basis for characterizing the heat transfer associated with fire during storage." This may or may not be conservative for the fire accident 

At the end paragraph (3) (c), reinsentences: concrete thfire is geneacceptable estimated odamage is 
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Comment SRP Location Public Comment Resolution Cha

accepted spalling of concrete due to fire without further evaluation. It also accepted concrete temperatures that exceeded ACI 349 limits as long as corrective actions are taken for continued safe storage. The revised version does not provide guidance on the structural assessment to fire event. Suggest restoring this paragraph. 

evaluation in a licensee's Part 72.212 site parameter report. To preserve the evaluation bases discussed in the original paragraph, only the lead sentence of the paragraph, which refers to Part 71 transportation provisions, will be deleted. 

and appropcorrective mpresumed. concrete teexceed the of ACI349 fproviding thresult from corrective abe taken fostorage. 

NEI 44 2962 Line 2962 states that 

consequences of floods such as damage to access routes, temporary blockage of ventilation passages, etc. “should be identified in the CoC so that a general licensee will be able to consider these factors when sitting an ISFSI”. This is a general site characterization issue more appropriate to be addressed in the 10 CFR 72.212 Report. Generic flooding depth and moving water limits the DSS is designed for should be described in the SAR and the CoC. 

Flood consequences, such as temporary but prolonged blockage of ventilation passages, may adversely affect thermal performance of the cask system. The staff agrees with the comment that evaluation of whether designbasis floods are bounded by floods analyzed in the certified cask system is a site-specific 72.212 characterization issue. Evaluation of additional flood consequences is generally at the discretion of cask vendors. 

Replace ththe word, “m

NEI 45 3083-3103 Lines 3085-3103 deal with the 

response of the storage system sitting on a flexible pad and subjected to earthquake loads. It requires that the 

The section of the document provides guidance to staff regarding the ISFSI seismic analysis and for reviewing calculations that show a cask will 

No change 
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Comment SRP Location Public Comment Resolution Cha

flexibility of the pad be taken into consideration in the seismic analysis. This is not an appropriate requirement for a system that is licensed to be used under a general license where the system design is based on a design response spectra (e.g. a RG 1.60 response spectra) anchored to a defined maximum acceleration for the horizontal directions and a maximum acceleration in the vertical direction. Each particular user is to ensure as part of their 72.212 evaluation that the system as qualified is adequate for each particular site considering the characteristics of the pad and its response when coupled with the underlying supporting media. 

not tipover or drop during a seismic event. 

NEI 46 3106, 12537 RG 1.60 imposes excessive 

conservatism for seismic evaluations. RG 1.60 should be replaced by NUREG/CR6728 and also NUREG/CR6865. 

RG 1.60 provides general guidance for generating design response spectra and has not been replaced by the cited NUREG reports. 

No change 

NEI 47 3139-3140 The term “confinement casks” 

is confusing. Should this be “confinement boundary”? 

Agree with comment Changed “c

to “the confthe cask”. 

NEI 48 3153 In the previous paragraph, 

Subsection NB is used to 

Agree with comment. To provide clarity, a sentence to recognize the 

Add a 2nd s

3.5.1.4 ii, 3
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define stress qualification for the confinement boundary, which is a pressure retaining boundary. In the paragraph including line 3153 it does not clearly state that the basket is a non pressure retaining boundary, and that the applicant should use Subsection NG. Need to state that Subsection NG is acceptable, or the reader is left to believe that Subsection NB applies to non pressure boundary baskets. It should also confirm that Appendix F is applicable for use with Subsection NG. 

code requirements for the basket is added. 

reads: “For Subsection applies.” 

NEI 49 3168 Although not a change from 

the existing version of NUREG 1536, this paragraph appears to imply that Section III analysis should be only linear elastic. This section should be clarified to allow elastic-plastic and other non-linear analysis as permitted by the Code. It should state that Subsection NB and Subsection NG do permit the use of Appendix F which does permit the use of inelastic properties for components which serve as the pressure boundary or also non-pressure boundary 

See response to NEI Comment 35 regarding use of inelastic material properties. The strain-based criteria are not recognized by the ASME Code or other applicable standards. Recognizing that the SRP provides review guidance for broad base, common subjects. An applicant, however, may propose, the NRC staff may consider the use of other acceptance criteria, such as strainbased criteria, only on a case by case basis with appropriate justification. The staff may review alternate strain-based proposals in 

See respon35 for the fcomment oproperties. No change strain-base
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applications, such as baskets. It should also state that strainbased criteria can be employed for energy-limited accident conditions, provided the applicant provides such basis for its use. 

greater depth depending on the applicability and experience with the criteria to the proposed DSS design. 

NEI 50 3171 In many applications for drop 

conditions, it should be acceptable to use strain-ratesensitive properties. Appendix F permits their use. Need to include “strain rate properties, which needs the appropriate references.” 

As worded, the SRP does not preclude use of strain-ratesensitive material properties for design analysis of cask drop conditions. 

No change 

NEI 51 3315 Editorial: Delete either “for” or 

“of.” 

Agree with the comment. Deleted the 

NEI 52 3321-3338 Please clarify the trunnion 

design stress criteria used to compare the stress at the trunnion connection with the cask body at that interface. Regarding Line 3338, “the applicant should evaluate the stresses and forces in the trunnion connections with cask body...”, since the cask body is typically designed per ASME Code Section NB, the NB stress criteria should be used instead of yield and tensile strength. Please clarify. 

The SRP provides guidance for implementing the ANSI/ANS N14.6 stress design factors evaluation by recognizing that the maximum bending stress occurs at the base of the trunnion. Implicit in this evaluation is a classical strength of materials approach to calculate the maximum average shear over the trunnion cross section. In the case of a loaded cask consisting of the transfer cask, as a special lifting device, and the loaded enclosure vessel in which the basket and its fuel assemblies are emplaced, the SRP provides that the applicant should evaluate the stresses and forces in the trunnion connections 

Changed ssentence in Section 3.5“If other asASME Secby the finite analysis anyielding at are consideshould provjustification
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Comment SRP Location Public Comment Resolution Cha

with the cask body and in the cask body near the trunnions. The Line 3336 statement will be revised to recognize potential localized materials yielding also in accordance with the ANSI/ANS N14.6 provisions for stress design factors. 

NEI 53 3380 Section 3.5.2 “Other System 

Components Important to Safety” does not contain the alternate concrete temperatures as listed in Lines 2612-2640 for the steel-lined concrete confinement cask structure. 

Agree with comment. The temperature limit alternative listed from 2607-2640 and 8196-8227 is added to Section 3.5.2 for consistency 

Incorporatereferenced the revision 39 into Sec

NEI 54 3747 “Appendix C” should read 

Appendix F for the version year of the ACI 349 that is described in Line 3501. 

Agree with comment Changed a

NEI 55 3758 Editorial: “30 ksi” should be “3 

ksi.” Also, should the example list include a maximum compressive strength because that value is a limit for drop and tipover analyses? 

Agree with comment. The NRC staff notes that the cask tipover analysis generally places a limit on the maximum concrete compressive strength of the cask storage pad. To ensure that the analyzed configuration remains applicable in a general licensee’s Part 72.212 site parameters evaluation report, an upper limit on the maximum compressive strength should be reported. 

In Section 3changed fifread, “3 ksirevise the eread, “Uppekgf/cm2) on strength of limit (3 ksi, concrete spstrength (f”on concrete analyzed aISFSI cask 

NEI 56 4182-4184/JS The sentence regarding 

delivery of electronic media is 

Agree with comment. Deleted se

noted that e
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Comment SRP Location Public Comment Resolution Cha

guidance for the applicant rather than the staff and as such should it may be more appropriate in another document. 

should be dappropriate if possible, sent to the Control Desduring secu

NEI 57 4302 The discussion about 

annotation of input files is too broad. It may be important for the reviewer to see and perhaps use the applicant’s files, but it is not necessary to understand all aspects of the input files. Some of these files come from Journal files or Log files which are generated by the program. It is not feasible to add comments to these files. Open ended statements such as adding “annotation” leads to vague expectations by the reviewer for the need of such documentation. 

Well documented input files expedite the review since it is easier for the reviewer to verify that analysis files are consistent with the design information provided in the SAR. If it is not feasible to add comments to some files, then, as indicated in the SRP, “the applicant should provide an adequate explanation of how computer models were assembled using the CMS in the appropriate SAR chapters or related documents.” 

No change 

NEI 58 4313-4315 Delete these lines. The level 

of review described here seems to be beyond an audit review and more like a third party validation of the computer analysis. It is the responsibility of the applicant’s QA program to ensure that the analyses are performed correctly. 

The level of review depends on the complexity of the application, including the uniqueness of new designs and safety margins. The guidance also reflects previous NRC licensing review experience in identifying insufficient analyses, (performed under applicant’s QA program) in these areas of computational analyses. 

No change 

NEI 59 4332 Clarify or delete “mesh type.” Agree with comment. Changed fi

sentences 
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Comment SRP Location Public Comment Resolution Cha

under Sens“The reviewthe applicasensitivity sCMS modeThis includmesh dens

NEI 60 4335-4336 A mesh sensitivity study is not 

required when stress linearization is being used for primary loading. Such detailed studies should be restricted to fatigue evaluations at stress discontinuities. 

A mesh sensitivity is required to make sure the analysis results are mesh independent. 

Added sentbullet of ApSensitivity sensitivity isure the anmesh indep

NEI 61 4349 Delete “plots.” Including plots 

of all results generates an 

enormous amount of unneeded data in the FSAR. 

As stated in the SRP, the SAR or related documents should include all relevant results (including plots). 

No change 

NEI 62 4411 The guidance stating that the 

decay heat removal system should operate reliably under off-normal and accident conditions is inappropriate given that some of the abnormal and accident conditions themselves involve impairment or loss of the decay heat removal system (e.g., blocked air ducts). 

Agree with comment. Revised SR

“Evidence mthe applicaheat removoperate reland loading 

NEI 63 4551 In item (2), it appears that this 

is an option addressing when fuel cladding temperature does

exceed 400oC (i.e., delete 

“not”). Please clarify. 

Agree with comment. For low burnup fuel the maximum allowable peak cladding temperature may be higher than 400°C as long as the hoop stress is less than 90 MPa, as indicated in the SRP. 

Review Re“(2) the matemperaturconditions 400°C (752
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NEI 64 4469-4471 Clarification should be 

provided for “address, quantify and report the degree of conservatism associated with the proposed models and the resulting safety margin.” This statement is vague. It is unclear what the specific information is requested and to what level of detail. 

Agree with comment Changed la

paragraph read: “The discuss, quthe SAR anassociated thermal modetail of the be comparathe SAR thanalytical thtable of resprovided in how the asconservatisparameters peak claddconfinemenetc.). The tbe supportedocumenteand calcula

NEI 65 4580 Editorial: Change “on” to “in.” Agree with comment. Changed aNEI 66 4612 Editorial: add a closing 

parentheses at the end of the sentence 

Agree with comment. Changed to 

technical spmaximum t(typically 46

NEI 67 4686-4687 Delete this sentence. It does 

not appear to add value to the review guidance. Alternatively, clarify why this is only applicable to horizontal basket designs. 

Partially agree with comment. Internal natural convection however should be verified through physical experiments or use of validated CFD codes. 

Changed laSection 4.5“Traditionamaintained convection should be vrobust CFD physical ex
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NEI 68 4768-4770 The SRP requires test data for 

each thermal effective conductivity. Are correlations from handbooks which are based on test data acceptable? Is test data still a requirement if a CFD submodel is used to calculate the effective conductivity as specified in Line 4686 to 4687? It is recommended that “from test data” be changed to “from test data, or CFD submodels, or other appropriate sources” 

Agree with comment. Changed 2

paragraph SRP to reathermal conthis manneshould verivalues havfrom test dasubmodelsappropriate representatgeometry, mtemperaturused in cur

NEI 69 4678-4681 Limiting convection to the 

outer surface of the cask contradicts already-approved designs that credit convection inside the fuel canister. This is clearly permissible with appropriate justification. 

Agree with comment. Deleted the 

from SRP: natural circlimited to thexternal suand the am

NEI 70 4687 Delete the word “robust.” 

Words like this are vague and subjective, allowing each reviewer to apply his or her personal definition of “robust” in their review and generate RAIs if the model is not “robust” enough. 

Partially agree with comment. This guidance is intended to advise the reviewer that CFD calculations are not trivial and many times are subjective to errors if not used adequately. 

Replaced tthe word “s

NEI 71 4742 Editorial: Delete misplaced 

closing parentheses in this line 

Agree with comment. Changed th

“width and channel.” Rafter “air chpunctuated 
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Comment SRP Location Public Comment Resolution Cha

commas. 

NEI 72 5041 Allowance should be made for 

a properly scaled mock-up instead of an “as-built cask system” to confirm the thermal analysis. 

Agree with comment. Design verification testing could be achieved by using as-built cask system or mock-up system 

Changed 1

paragraph read: “As aconfirmatorapplicant mperform detesting of aproperly scsystem (whconfirm the presented i

NEI 73 5185-5187 Delete or clarify this sentence. 

No such “periodic surveillance program” has “typically” been required or performed for stainless steel welded canister confinement systems. Periodic surveillance of the confinement boundary, if any, should only be required casespecifically, if the particular design features of the confinement system require it. Inspections of the air vents or temperature monitoring have been accepted as the sole periodic surveillance. 

Agree with comment. Replaced s

the followinconsistent wother weldeconfinemenmonitored sreview ensthe confinethe licensinsurveillancefor blockagtemperaturdepending design.” 

NEI 74 5347-5348 The statement that the 

monitoring systems are not important to safety and

classified as Category B (an ITS class) does not appear to be consistent. 

Monitoring systems are not specifically mentioned in NUREG/CR-6407 “Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety”. ISG-5 which 

Added “It is important to of the assohave not msafety proglike design, the 3rd para
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Comment SRP Location Public Comment Resolution Cha

was incorporated into the revised SRP, mentions that monitoring systems are a Classification Category B because as stated in Table 2, a Category B component is one whose failure in conjunction with the failure of an additional item, like the containment boundary seal, could result in an unsafe condition (potential release of radioactive material). It is termed as not important to safety since most of the associated hardware have not met the important to safety programmatic controls, like design, or procurement. 

last senten

NEI 75 5384 Editorial: Change “Review” to 

“Evaluation.” 

Agree with comment Changed a

NEI 76 5413-5426 This paragraph does not 

appear to be consistent with ISG-5 and ISG-18 and would only apply to non-weldedcanister type confinement systems. Based on NUREG/CR-6397, damaged fuel would not have a driving force to release fines form from the fuel matrix. What is the technical or safety issue of concern? What factors are

suggested for damaged fuel? 

If the canister is welded and tested to be leaktight then the size of the source term is immaterial for determining a release. The staff agrees that fuel rods that are classified as damaged due to a preloading cladding breach may not have a pressurized fuel rod driving force for the release of particulates from the rod. However, under an impact accident, damaged fuel rods might release additional fuel fines due to the fracture of the fuel, especially in the rim region of high burnup fuel. In addition, some canisters may be 

Added to e

paragraph Fuel rods thdamaged dcladding bra pressurizforce for thparticulate off-normal basis eventan impact afuel rods madditional ffracture of tthe in the riburnup fue
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pressurized to several atmospheres and cask blowdown may also affect release fractions. Each applicant should establish release fractions for damaged fuel based on applicable physical data and other analyses appropriate for the specific type of fuel, damaged condition, and accident conditions. This will be clarified in Section 5.5.3. Alternatively, a leak-tight confinement boundary may be specified to preclude the release analyses of damaged fuel. Also, see resolution to NAC 5426. 

canisters mseveral atmblowdown creleases fraapplicant srelease fracfuel based physical daanalyses aspecific typimpacts, ancondition oa leak-tight boundary mpreclude thof damaged 

NEI 77 5800-5801 "radionuclide content, and 

estimated radiation source strength in Becquerel's, .... should be described": This appears to be a new expectation from the NRC. It is not clear what the basis of this request is as radiation source strength in Ci or Bq is not clearly related to gamma/neutron source strength (e.g. beta emitters). 

This guidance is provided to evaluate source terms of different types of contents, for both the shielding and confinement analyses. The SRP is revised toclarify this guidance.

Replaced tSection 6.4“The physicform, ...” w“For spent source termMeV/s per described igroup strucfunction of fuel hardwaor Becquercontents otfuel hardwaisotopic comphoton yiel
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Comment SRP Location Public Comment Resolution Cha

constituent For confinepurposes, tchemical fogeometry, rand estimastrength sh

NEI 78 5809-5810 "characteristics for each 

gamma-ray source type should be provided, including isotopic composition, and photon yields”: Is a tabulation of spent fuel isotopics requested here? If so, for what purpose? Typically, inputs into depletion analysis are provided, but not isotopics of depleted materials. 

The guidance is intended for different types of contents in the shielding and areas of review. Isotopic concentrations are needed because the input file, as described in the SRP, is typically a representative input and not necessarily bounding. For cases in which the source terms are not derived from a depletion calculation, the applicant should provide isotopic concentration and photon yields. The SRP is revised to clarify this guidance.

Replaced t

Section 6.4following: s“The SAR sgamma souspent fuel amaterials. fuels, the sbe describecompatible calculation form of phoenergy bin. hardware ahardware, sbe specifie

Curies or Bcontents otfuel hardwaisotopic comphoton yielconstituent A tabulated radiologicaacceptable

NEI 79 5813-5814 Within gamma source Agree with comment Replaced w
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Comment SRP Location Public Comment Resolution Cha

description "describe extent to which radioactivity may be induced by interactions involving neutron originating in the stored materials": If this implies n-gamma reactions, then the current SRP version is clearer. If activation is to be considered for decommissioning, that should be clarified. 

statement: “The SAR sdiscussion radiations creactions spackaging contents.” 

NEI 80 5868-5870 Shielding analyses do not 

need to be “bounding analyses.” Applicants need only provide representative dose rates to demonstrate reasonable assurance that the system is capable of meeting the offsite dose limits or 72.104 for an entire ISFSI. (See line 5723 and subsequent text.) 

Shielding analyses should provide the bounding dose rates and demonstrate that the system is capable of meeting the requirements of 72.104 and 72.106. The bounding doses rates should be based on the design basis loadings that are defined through the applicant by maximum burnups, minimum cooling times, and minimum enrichments. 

No Change 

NEI 81 5873-5882 High burnup fuel has been 

licensed for storage on several dockets. There is no indication that high burnup fuel produces substantially high dose rates due to limited validation data. If limited data is available it leaves an open ended question as to how to specify uncertainties. “Conservative assumptions” and “design margins” are not defined, leaving it up to each reviewer 

Fuel assembly with higher burnup will produce higher gamma and neutron sources. The gamma source increases linearly proportional to burnup and the neutron source increases proportional to the fourth power of burnup. It is expected that the magnitude of uncertainties in exposure rates and decay heats would propagate proportionally in the same manner. There exist biases and 
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when, and how much, in uncertainties to apply. There is no correlation as to how maximum fuel assembly heat load is related to uncertainties - low heat capacity /minimal shield system may be affected by low fuel assembly heat load, and vice versa. 

uncertainties in the computer model and input data. These errors, bias, and uncertainties are in general not quantified if the computer code and models are not benchmarked and validated against experimental data. The NRC recognizes that the nuclear industry has not developed experimental data for the high burnup fuel that is proposed for storage. NRC has traditionally allowed applicants to use isotopic codes beyond their validated range. However, some applicants have applied penalties to assure these un-quantified uncertainties are sufficiently considered for high burnup fuel source term and decay heat predictions. The penalty factors should account for reasonable uncertainties in both gamma radiation, neutron radiation, and decay heat source terms. The magnitude of uncertainties in these three source terms may be significantly different at high burnups. Alternatively, the applicant may propose measurement programs in the technical specifications that directly monitor shielding and thermal performance (e.g., cladding temperatures,, and detect 
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abnormalities that could result from unaccounted uncertainties in the source term predictions. Regarding the uniformity of the practice among the staff on the additional safety margin, the Criticality Safety and Dose Assessment branch of SFST has working groups to share review experience and develop consensus. The staff is in general aware of the common practice. 

NEI 82 5968 Editorial: Incorrect spelling of 

“Principle.” 

Agree with comment Changed a

NEI 83 5996 Editorial: Figure 6-2 is missing 

from the document. 

Agree with comment In Section 6

words: “in F(reproduce6716)” withThe larger generates a content whneutron sousecondary as illustrate6716, Sect

NEI 84 6003-6004 "...applicant and the staff 

should not attempt to establish specific source terms as operating control and limits for cask use.": If this is true, why does the SRP focus in the Section 6.4.2 on curie content and isotopic description of the 

The focus of this requirement is different from that of Section 6.4.2. The requirement here is for consideration of the cask operations. Fuel assembly initial enrichment, burnup, and cooling time are the readily usable and inspectable parameters for cask 

Replaced laSection 6.5“However, limits for caHowever, tattempt to uterms as ba
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Comment SRP Location Public Comment Resolution Cha

spent fuel? For Cobalt-60 dominated hardware sources, a source term may be more appropriate than other limits (e.g., mass, exposure, cool time). 

operation. Source terms would be additional parameters that are calculated from enrichment, burnup, and cooling time. 

operating ccask use breadily inspThe fuel asenrichmentcooling timappropriate controls an

NEI 85 6036 Editorial: add a closing 

parenthesis. 

Agree with comment Changed a

NEI 86 6449-6450 "...homogenization should not 

be used in neutron dose calculation when significant neutron multiplication can result from moderated neutrons...": While not changed from the current SRP, it should be noted that standard, NRC-approved, practice is to homogenize the rod lattice in shielding calculations (not necessarily homogenizing basket structure into the fuel region). 

Although this assumption has been acceptable in many applications, there may be instances where homogenization may not be appropriate. 

No change 

NEI 87 6188 Incorrect spelling of the word 

“Evaluation” 

Agree with comment Changed a

NEI 88 6221-6222 Review staff should recognize 

that importance functions may also be produced with Monte Carlo, point-kernel and transport codes. 

Agree with comment. Replaced t

the 2nd para

6.5.4.1 with “The reviewthat the apptransport omethods to and/or gamfunctions (u
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Comment SRP Location Public Comment Resolution Cha

mrem/hr/paAdded the to the end o

in Section 6“The reviewpay close aapplicability function to content and contents an

NEI 89 6246-6248 "The applicant should use the 

latest released computer code version that is valid for the particular computational platform used to perform the analysis.": This item in particular has been discussed with NRC staff as a significant issue. A licensed code for the same type of application should not require a code version change simply because the code developer has issued a new version. Use of different code versions within one or more applications is difficult to reconcile and potentially leads to unnecessary confusion. Such burdens should only be borne by the applicant if a significant safety issue has been identified with the previous code version. Typical 

Partially agree with comment. The staff would prefer models to be based on latest released computer code versions because NRC typically upgrades its shielding computer codes on a regular schedule with code vendor upgrades. However, computer codes used for shielding analyses do not necessarily need to be updated to the most recent version. The applicant should demonstrate that use of a code version, that is no longer supported by a vendor, is valid for the specific analysis, and also that the code has been properly maintained in accordance with the requirements contained in 10 CFR Part 72, “Subpart GQuality Assurance.” The SRP is revised to clarify this guidance. The letter, dated July 2, 2009, from Mr. Raymond Lorson to Mr. Steven 

Replaced t“The appliccomputer cdemonstratfor the anathe particulplatform usanalysis. Tconsider if confirmatorreview is neshielding papplicant thunsupportein cases weupgraded chave the caexamine unmodels from 
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new release code versions tend to contain a certain amount of bugs that get resolved through user feedback to code originator. While it could be postulated that newer code provide more "accurate“ results, but if the previous version was found to be acceptable for system approval with no safety issues identified, why should applicants be required to change? The goal per draft SRP Section 6.4 is to provide reasonable assurance that system will meet limits. This is also inconsistent with how NRR deals with updated codes (e.g., ASME Code). 

P. Kraft (ADAMS: ML0918802633) provides the regulatory basis and detailed explanations for this requirement. The applicant’s quality assurance program should also be capable of identifying and addressing “bugs” in cases in which they chose to use new codes for their shielding analyses. 

NEI 90 6302-6309 "by verifying that the following 

information has been provided in the SAR ... The computer code solutions to a series of test problems ...": The draft SRP revision does not contain the previous SRP statement "that these solutions may be referenced, and need not be submitted in the SAR". This change would add a substantial amount of information to the SAR without any safety benefit as the referenced documents, per 

Agree with comment Added to th

following wspecificatioavailable recommonly established and MCNPvalidation.
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Comment SRP Location Public Comment Resolution Cha

current SRP, should be public information and/or have been previously submitted to NRC. 

NEI 91 6578 This implies that only boron 

can be employed as a fixed absorber. It is recommended that “boron” be changed to “neutron poison material” 

Agree with comment. Changed to 

poison mat

NEI 92 6739 Neither Section 8.5.4.3 nor 

Attachment 8-3 exists in the document. 

Agree with comment. The citation

the SRP ar

NEI 93 7099-7104 This section requires explicit 

analyses of atypical control rod insertion while Section 7.5.5.6 (lines 7138-7157) discusses margin to cover higher-thanmodeled reactivity due to control rod insertion. These two sections appear to conflict. Please clarify what is required in the design basis calculations. 

Agree with comment. The followi

inserted in “While the make every and appropthese potenexplicitly, dmake it diffthese unceand assure adequately 

NEI 94 7102-7104 These lines explicitly require 

the analysis of integral fuel burnable absorbers. However, there are NUREG/CR reports that provide guidance on when these absorbers need to be considered in the analysis. These lines should be revised accordingly. 

Agree with comment. A reference 

6760 is addof Section 7

NEI 95 7242 

7390 

 “Foreign standards are not generally acceptable...” What is the basis for this statement? For non-ASME code 

Agree with comment. The applicant must provide an analysis that shows the foreign standard is equivalent to a comparable US 

Changed w8.1, 3rd par

8.4.2.1, 3rd 
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Comment SRP Location Public Comment Resolution Cha

applications, there are many recognized standards essentially equivalent to ASTM, such as Euronorm, JIS, etc. The applicant should be able to use foreign standards with appropriate justification. 

standard, or otherwise sufficient for its intended use. The staff may review foreign standards in greater depth depending on the familiarity and applicability of the standard to the proposed DSS design. 

Foreign stamay be accby-case bashould provdocumentause of the fshow that tis equivalenUS standarASTM, etc.sufficient foThe staff mforeign standepth, depefamiliarity wapplicability the propose

NEI 96 7248 The Chapter 8 convention of 

indicating with an asterisk the items that should be addressed in the Technical Specifications is not used in any other chapter. All of the chapters should be consistent and not use this convention. 

Agree with comment. The convention is removed. The specification of review areas that should be considered for Technical Specifications is clarified in the text of Chapters 8 and 13. 

Section 8.4revised to cshould be ctechnical sp

NEI 97 7266 

7554-7564 

Replace “weathering steel” with “0.20% copper steel” or “carbon steel with a minimum copper content of 0.20%”. Also, add “salt water” to “coastal marine sites”. The term “weathering steel” applies to a class of low-alloy steels that contain small amounts of such alloying elements as Cr, 

Agree with comment Changed S

To address atmospherifound at cowater) siteshave specif0.20%, minbearing stesteels” suc
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Comment SRP Location Public Comment Resolution Cha

Ni, P, Si and Cu. These steels are covered by ASTM A242 and A588. Also “copper bearing steel” should be generalized to allow for other appropriate measures to control corrosion. 

Kennedy Shas collectedemonstratcopper-beasteels for scorrosion asites. Thermarine ISFcopper-bea(containing percent copsteels, may Such steels ASTM A-24supplemenASTM A-36specificatioOther corromeasures mprovided addocumentademonstrat

NEI 98 7317-7321 This paragraph should be 

deleted for several reasons. The portion of the sentence stating that the body of the SAR “is not enforceable” is incorrect. Users must comply with the Part 72 cask SAR unless a change, appropriately reviewed and authorized under the provisions of 10 CFR 72.48, is performed. If not, NRC enforcement action may 

Partially agree with the comment. General licensees and cask vendors have authority to change an FSAR under the requirements of 10 CFR 72.48. If 10 CFR 72.48 is not performed correctly, the NRC may take enforcement action. Enforcement action may only be taken for violation of regulatory requirements, license/CoC 

Also see reThis paragrHowever, tremoved bebelong in SProcedures Criteria” of chapter. Thmore approChapter. 
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be taken. In addition, using this logic as the basis for putting information in the CoC or TS is flawed because it is not risk-informed, is too subjective, and dilutes the CoC holder’s and licensee’s ability to implement changes that meet the criteria of §72.48. Moreover, this increases the NRC’s need to spend resources reviewing changes to the CoC that are not risk- or safety-significant. 

(including TS which are appendices to the CoC) conditions, and NRC Orders. Paragraph 72.48(c)(1)(B) recognizes the role of certificate conditions and TS in limiting design changes without NRC approval. These certificate conditions and TS are established at the discretion of NRC for design features, operations, and contents that should not be changed without additional NRC licensing review .. 

NEI 99 7334 a) Amendments are not 

“completely new designs.” New designs are submitted as a new CoCs. This statement should be revised. b) Use of the term “beware” is derogatory in that it implies the applicants are trying to sneak changes through the NRC without them being noticed. Please revise. 

a) Although some modified designs have been submitted for NRC review as new certificate applications, other modified designs which represent new, major design components such as canisters, storage overpacks, and transfer casks have been submitted as CoC amendment requests. The staff will revise the phrase to clarify this issue. b) The statement is not intended to be derogatory towards vendors. Given past review experience, each vendor has used unique styles and formats in their amendment requests, including the integration of new analyses into existing FSARs, and the demarcation of textual changes 

Section 8.4begin as foshould survdesign drawSAR Chaptidentify the issues that with the spproposal in The reviewexamine thshielding, cthermal chacross-cuttinbe coordinatechnical dall design ain the amennecessarily separately applicant n
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and 72.48 changes. It has been challenging for the staff in some cases to understand exactly what information is new and has changed since the last version of the FSAR was formally reviewed by staff in a previous licensing actions. In some cases, new or removed information has not been properly identified, or was ambiguous to the staff, in SAR change pages submitted during the review process. Given this experience, the purpose of the statement is meant to caution the reviewer to not overlook all potential changes in an amendment request, and ensure there is a clear understanding of the changes being requested for approval. However, the statement is revised for clarity and moved to the Introduction of the SRP for generic application to all disciplines. 

The reviewthe followinSpecificatioverify its prapplicant aspecific limrequiremenconstraints applicant.” 

NEI 100 7338-7345 This paragraph should be 

deleted for a couple of reasons. It is incorrect to state that things previously approved and outside the scope of the amendment request are subject to review again. This is contrary to good regulatory practice and rereviewing approved 

Amendments such as content and design changes, are founded upon the design and methodologies previously reviewed by NRC. Compliance of a DSS is often based on the performance of the contents, canister, and overpack as a system. As a result, portions of these designs and methodologies in the SAR may be re-examined as 

Removed tparagraph following to the fifth parReview ProSome amecontent and are foundeand method
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information could create a contradiction with a previous staff SER. In addition, the sentence in lines 7341 and 7342 could be viewed as derogatory towards both the NRC project management and the applicant. 

part of good regulatory practice to ensure the new amendment proposal meets Part 72 requirements. It is not the intent of the staff to reexamine designs previously approved in a CoC for re-approval. However, an amendment audit review may from time to time detect deficiencies or errors that were not identified during a previous audit review. It should be noted that it is the primary responsibility of the cask vendor to ensure such errors do not exist. Also, new information regarding operational experience or new phenomena may come to light which requires NRC consideration, in order to assure the design remains safe and compliant with applicable regulations. However, the statement is revised to clarify this issue and is moved to the Introduction of the SRP for generic application to all disciplines. Issues involving the licensing process (line 7341) and interactions are appropriately described in internal operating procedures for NRC staff. Therefore, this discussion is 

reviewed bsystem. Evamendmenare often bperformanccanister, anintegrated sportions of componentmethodologbe re-examthe new sysamendmenPart 72 reqthe audit reamendmenoccasionallsafety quesof the previdesign. The review that ask questiodesign remcompliant wregulationsshould be lunderstandthe specific should conprecedentsguidance, aas approprshould also processes enforcemen
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eliminated from the SRP. issue progr

these potenquestions wapproved d

NEI 101 7362-7263 “copper bearing structural 

carbon steel” should be generalized to allow for other appropriate measures to control corrosion. Also, it seems inappropriate to single out one DSS design in review guidance. 

Agree with comment Changed th

8.4.1 to: “Uor weatheristructural scoastal maother corromeasures)

NEI 102 7382 This should read “All ASME 

materials are a subset of AWS and ASTM materials” 

Agree with comment Changed a

NEI 103 7394 The statement that all ITS 

materials are typically ASME II materials is not correct. That is only true of components subject to ASME Section III jurisdiction, typically confinement boundary and fuel basket. ITS attachments to the confinement boundary, as well as structural components of the overpack, are likely not ASME section II materials; for non-ASME ITS components, ASTM materials can be used. 

Agree with comment Changed 4

Section 8.4ITS compoASME Sectypically coand fuel baASME Secattachmentboundary, acomponentmay be ASmaterials, dcode of reccomponentITS compomaterials m

NEI 104 7400 Non-ITS materials specified to 

ASTM. This is not correct. According to Reg Guide 7.10, Appendix A, ITS Category B 

Agree with comment Changed 5

Section 8.4Non-ITS ite
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must be used in accordance with rigorous specifications; ITS Category C need not. Therefore, it is correct to state that ITS A and B should be specified to ASTM, ASME, or equivalent standards; ITS Category C, and non-ITS items can be specified by generic names such as “stainless steel”, “aluminum,” “carbon steel,” etc., as appropriate for the application. 

by generic “stainless s“carbon steappropriate 

NEI 105 7408 Editorial: Delete. This line 

repeats lines 7391-7392. 

Agree with comment Changed a

NEI 106 7411-7412 No changes in neutron 

absorbers without NRC review. This is not correct; changes should be acceptable with appropriate review or testing by the certificate holder, with only select critical limiting characteristics included in the TS. 72.48 provides adequate change control for these items given the risk of dry cask storage operations. 

Agree with comment Changed 6

Section 8.4Proprietary ITS (specifpoisons) madequately “Materials” make a safgoverning qand quality documentsproceduresprotocols fomaterials sincorporatethe TS. Limmaterials cperformancmethods m
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changes sa10 CFR 72

NEI 107 7420-7425 Editorial: This information 

repeats prior information. 

Agree with comment Deleted su

NEI 108 7470-7471 Remove “transportation” as 

transfer is already listed. Remove “retrieval”. In this context it is the same as unloading. 

All of the terms in this sentence are applicable to the storage activity except “transportation” 

Removed “the sentenc

 NEI 109 7515-7518 The information pertaining to 

steel producers is unnecessary for review guidance and should be deleted. If it is retained, at a minimum delete the last sentence regarding “defeating” a steel producer and clarify who is meant by “steel producers.” 

Agree with comment. The information pertaining to steel producers was meant to reflect lessons learned in past evaluations of steel certification. However, the language is clarified to only reflect the use of ASME Code values and CMTR values. 

Replaced t6th paragra

with: “Examadopted maITS compoensure ASMD, propertieemployed. (developed material proASME Codused. Use test report UTS, yield, permissiblevalues is at conservativmay be takingot, billetoptimum mduring certi

NEI 110 7520-7523 This paragraph appears to be 

an editorial opinion and serves no value as review guidance. Delete. 

Agree with comment. Paragraph 

NEI 111 7554-7557 References to specific dry Agree with comment. Revised to 
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storage vendors are typically not appropriate in the SRP. Please consider revising this section. If reference to a vendor is appropriate, the corporate name should be used rather than abbreviations. Therefore, change TN to Transnuclear, Inc. 

names 

NEI 112 7562-7564 What is the basis for no credit 

for coatings unless periodically inspected? Thermal spray AlZn coatings and hot dip galvanizing are widely used in marine applications, and are much more predictable than paint with respect to adhesion. 

Without supporting data to demonstrate predicted coating life, monitoring is needed to assure intended performance. The guidance is revised to clarify this information. 

Changed laSection 8.4Coatings malleviate thatmospheriThe coating periodically maintaineddata is avaa predicted 

NEI 113 7577 It is recommended that “AWS 

D1.6 (current edition), “Structural Welding Code – Stainless Steel” be added to this list of codes. 

Agree with comment Add “AWS 

edition) StrCode-StainSection 8.4Codes. 

NEI 114 7608 The full penetration welds 

should only apply to the confinement boundary of the canister. In some designs the bottom closure weld is not a confinement boundary weld.  For non-confinement boundary welds, other design should be acceptable. Please 

Agree with comment Changed fi

Section 8.4the canisteare full pen
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clarify 

NEI 115 7621-7622 

8465 

“helium leakage test is performed of the entire shell” – Please clarify that this testing only applies to the confinement pressure boundary (i.e., not attachment shell welds). 

Agree with comment, except that Line 8465 is not applicable to this comment. The NRC has also issued draft ISG-25 “Pressure Testing of Confinement Boundaries” This guidance has been administratively incorporated into the SRP, which addresses this comment. 

Revise Secincorporate guidance o

NEI 116 7621-7622 What is the basis for requiring 

a helium leakage test? The confinement boundary is designed in accordance with ASME Section NB, NC, or ND. The Code includes pressure tests to confirm pressure boundary integrity. If this is sufficient for high pressure vessels and piping systems in a power plant, it should be acceptable for a confinement boundary given the relative risk and service conditions. 

See response to NEI Comment 115. Pressure tests, examinations, and leakage tests serve different functions. The volumetric and surface examinations of welds ensure geometric compatibility with the design requirements, but can only detect flaws down to a certain size. The ASME Code pressure test provides additional assurance that the component has been properly fabricated by stressing the component to a minimum Code required loading. The helium leakage test ensures there are no flaws or leak paths that could result in significant release of the helium and radioactive content to the environment. The weld nondestructive examinations, ASME Code pressure test, and helium leakage test are not considered 

See respon115. 
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equivalent substitutes for each other. The regulations mentioned in the text (i.e. 72.236 (d), (j), and (l)) provide the regulatory basis for the helium leakage rate test. Designing a component in accordance with ASME Code does not ensure that it is fabricated to prevent small potential gaseous leaks. Section 8.4.7.2 is updated to clarify guidance for leakage tests and to administratively incorporate the guidance of draft ISG-25. Sections 10.5.1.1 and 10.5.1.2 also capture this guidance. 

NEI 117 7624-7625 Not all of these tests (e.g., 

hydrostatic or pneumatic) are performed in the fabrication shop. Testing is in accordance with the design code. No additional review guidance is necessary. Shop helium testing would be an additional commitment beyond what the design code requires. Please clarify. 

The helium leakage test provides assurance there are no flaws or leak paths that could result in significant release of the helium and radioactive contents to the environment. It is required to demonstrate compliance with 10CFR 72.236 (d), (j), & (l).The Code required pressure test ensures fabrication integrity of the component, but it does not ensure prevention of small gas leaks. Meeting Code requirements for pressure testing does not ensure meeting regulatory requirements for helium leakage rate testing. He leakage testing derives from Part 72, not ASME. 

Section 8.4updated to guidance oto response 115 and 11
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Also refer to Sections 10.5.1.1 and 10.5.1.2 that capture this guidance for pressure testing and leak testing, respectively. 

NEI 118 7630 Editorial: Add “as” after “or.” Agree with comment Changed aNEI 119 7641 Editorial: Change “designedto” 

to “designed to.” 

Agree with comment Changed a

NEI 120 7646 The N45.2 series has been 

replaced by NQA-1. Suggest referring to both for older commitments and newer commitments to the QA code. 

Agree with comment Changed th

the fourth b8.4.7.3 to: documentincomply with 10 CFR PaAssurance NQA-1, “QuRequiremeFacility Appdepending effect at the 

NEI 121 7697-7701 For stainless steel canisters 

and welding, this is too limiting. The J-integral method to evaluation flaw size is used, which limits the size of a single weld pass. In order to be consistent with line 7682, it should explicitly state that the applicant can use J-integral methodology incorporating plasticity for ductile weld materials such as stainless steel. 

Agree with Comment Revised Se

identify the ultrasonic openetrant estructural liGuidance idetermining including usnet section 

NEI 122 7700 The canister is designed per 

ASME Section III, Division 1, 

Agree with comment. Division 3 has not yet been endorsed by 

Changed D“Division 1,
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Subsection NB, not Division 3. Has Division 3 been endorsed by NRC? If so, both Division 3 and Division 1 should be discussed. If not, reference to Division 3 should be deleted. 

NRC. 

NEI 123 7715 Delete “Pursuant to NRC to 

Bulletin 96-04 (1996).” This language implies regulatory requirements are contained in the bulletin. An NRC bulletin is a request for information at a particular point in time. It is not something to be referenced as a source of information upon which to base a review of an application. The SRP should stand alone and refer to regulations and approved guidance only. 

Partially agree with comment. An NRC Bulletin is not a regulatory requirement. Bulletin 96-04 addressees the potential for chemical, galvanic, or other reactions among the materials of a spent fuel storage cask, to assure no adverse reactions exist. This guidance is still applicable to the certification of DSS in order to meet the regulatory requirements of 10 CFR Part 72. The SRP is revised to clarify that the Bulletin may be used for guidance. 

Changed to can find opassociated generation evaluating reactions in (1996). Alsshould conperform adoperating eexpected (eloading/unlstorage) folicense peradverse gareactions ocanister maand the opeenvironmen

NEI 124 7743 The statement that aluminumbased metal matrix 

composites are employed for all presently utilized neutron poison materials is incorrect. Boral, for example, is used through the industry and is not a metal-matrix composite. 

Agree with comment Replaced w

based metaand aluminlaminates (

employed futilized neumaterials. 

NEI 125 7750 

7763 

Analysis of creep for all aluminum based structural 

Gap analysis can change drastically if aluminum components 

Changed 1

paragraph 
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materials, including those only supporting dead weight – “any kind of loading.” There is no sound basis for requiring a creep review of materials that have no structural function except bearing accident loads through their thickness, and supporting their own dead weight during normal storage. 

creep and increase basket gap. Review the 

temperaturaluminum cverify a crebeen perfoload bearincomponentdesign temapproximat

NEI 126 7724 

7824 7881 

This section is entitled “Exterior Protective Coatings” but lines 7824 and 7881 refer to interior coatings. 

Agree with comment. The section is applicable to both interior and exterior coatings. 

Title of Secrevised to tCoatings. 

NEI 127 7772 Exterior coatings. Scope and 

level of review for this area appears excessive and inconsistent with the “low priority” given. This should be reduced to specifying the generic coating systems that are acceptable, with surface preparation and paint application in accordance with manufacturer’s instructions. Specifying the manufacturer and submitting the paint technical data sheets requiring qualification testing (lines 7881) are overly burdensome given the low risk. 

Partially agree with comment. With the exception of coating issues that may result in adverse chemical or galvanic reactions described in NRC Bulletin 96-04, coatings are generally a low priority item with low safety significance. In these instances, most of the guidance in this section is not applicable. However, instances may exist in which unique or innovative coatings are specified by the applicant to perform a specific function unique to the cask system. In these instances, the reviewer may use discretion in implementing the detailed guidance in this section. 

Section 8.4include “Cohave a low with the exissues that adverse chreactions. detailed gusection is nto further cof the reviemay be insunique or inare specifieperform a sunique to ththese instamay use diimplementireview guid

NEI 128 7824-7825 It is not necessary to include See response to NEI 127. Most See respon
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the coating manufacturer’s technical literature in the SAR. The critical characteristics of the coating material are what is important and should be sufficient. The supplier should be free to use whatever coating material and manufacturer that has these characteristics for the service conditions. 

coatings have unique properties and application steps. Often the characteristics of the coating and coating performance are dependent on the precise steps that were taken to apply the coating. 

Deleted “Thmanufacturliterature fospecified fobe submittestaff reviewunique natuproperties, application manufacturbe the only information coating. 

NEI 129 7832-7942 Delete Sections 8.4.11.4 

through 8.4.11.6. Surface preparation coating repairs, and coating qualification testing are all details not necessary for the staff to review. These attributes of the coating system are dictated by the coating manufacturer or the CoC holder for the particular coating material and service conditions. Appropriate surface preparation, repairs and qualification testing are all adequately governed by the CoC holder’s or licensee’s coating specification and procedures developed under the applicable QA program and the coating manufacturer’s requirements. All of the above is subject to NRC inspection 

See response to NEI 127 See respon
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for verification of compliance. 

NEI 130 7882-7884 It appears that this sentence is 

written for paints and does not account for the possibility of plating as a coating. 

The statements in this paragraph are applicable to any coating, including paints or plating. 

Phrase “(inplating)” wasentence. 

NEI 131 7950 The statement that neutron 

shielding materials are not ITS appears to conflict with NUREG/CR-6407, which specifies that shielding materials are ITS Category B. Please clarify. 

Agee with comment Paragraph 

Shielding Mrevised to ishielding mStaff also nqualification testing of nmaterials srequired in characteristo performacompositioneutron shishould be s

NEI 132 7963 The first sentence in this line is 

unnecessary. Delete. 

Agree with comment Changed a

NEI 133 8021 Impurity limits may or may not 

be established as a result of qualification testing; that is not the main purpose of qualification testing. 

Agree with comment Deleted the 

“Qualificatiouseful in esimpurity coborated abexceeded. limits can bagreement seller.” 

NEI 134 8008 Editorial: “Surrey” should be 

“Surry.” 

Agree with comment Changed a

NEI 135 8048 Submittal of manufacturer’s 

data sheet for neutron 

Agree with comment Replaced; 

data sheet 
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Comment SRP Location Public Comment Resolution Cha

absorber is only applicable if the applicant is proposing a trade name product. Add “as applicable” at the end of the sentence. 

to suppleminformation“If the applian absorbespecific tramanufacturshould be ssupplemeninformation

NEI 136 8103 ZrB2 standard: All standards 

are a compromise of some kind: homogeneous standards like ZrB2 must be paired with aluminum sheets to simulate the scattering by aluminum in the neutron absorber; scattering by carbon in boron carbide is generally not simulated. Non-homogeneous standards that have a very fine uniform dispersion of the boron-containing phase are only an approximation of the homogeneous material assumed in the criticality safety calculations, but they get the appropriate aluminum and carbon scattering. Therefore, change “a qualified homogeneous standard such as ZrB2” to “a calibrated standard that is either homogeneous, such as ZrB2, or that has a very fine and 

Agree with comment Changed to

Aa = accepneutron attqualified hoabsorber stZrB2, or a hcalibration traceable to recognized calibrated wneutron becross sectioCalibration evaluated a1/0.90) of tassumed in computatio
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uniform dispersion of boron such that it approximates homogeneity.” 

NEI 137 8110 P=0.999: Previously the staff 

has accepted P=0.95 and should continue to do so considering all the conservatisms involved (e.g. keff ≤ 0.95, the 90% maximum credit for boron 10). 

Agree with comment Changed; 

“Let P = 0.9

To: “Let P = 0.9

NEI 138 8122 Quantitative measures 

(porosity testing, tensile testing, etc.) are now preferred over qualitative examination (TEM, SEM). Metallic/ceramic systems are generally accepted as not susceptible to radiation damage from gammas or from neutrons at the fluences encountered in dry storage. 

Agree with comment . 

Replaced fof Section 8detailed dequalification accepted badded qualneeded to bstructural n

NEI 139 8155 A sample from every other 

piece is too prescriptive for a standard review plan; according to ASTM C1671, random or systematic sampling should be applied. 

Agree with comment Changed: 

“Adequate should be tother comp“Adequate should be tcomponent

NEI 140 8156-8157 Lot definition based on billet 

may not be appropriate for material from small billets; allow alternate definitions that are uniform for sampling purposes. 

Agree with comment Removed 

defining ‘lot

NEI 141 8186 Please delete the following ACI 349 Section R3.5.3 States: Changed th
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sentence “Zinc, zinc rich coatings, zinc-clad materials, and aluminum should not be used for any embedded objects that will be in contact with wet concrete, because of the potential for concrete degradation from an adverse chemical reaction”. Zinc galvanized reinforcing steel and zinc plated/galvanized embedded lifting devices are common and widely used in the concrete industry. Even though chemical reaction between the zinc and water in concrete may occur at any age, this reaction is not proven to have any adverse impact on concrete. Note that Section 3.5.3.8 of ACI 318-08 allows the use of galvanized reinforcing steel per ASTM A 767. 

“Deformed reinforcement—Zinc used in the galvanizing process may negatively react with alkaline materials commonly found in concrete. In addition, potential galvanic corrosion with other embedded metals, as well as hydrogen generation and potential for hydrogen embrittlement, suggest that such coatings may be detrimental. Research conducted by Sergi et al.3.1 concluded that zinc coatings provide little value in providing long-term protection of reinforcing steel, and cautionary statements in ACI 201.2R3.2 support this position. These industry concerns have prompted ACI Committee 349 to prohibit the use of zinc coatings on reinforcing steel in nuclear safetyrelated structures until adequate data justifying its use can be reviewed.” 

“Zinc, zinc clad materishould not embedded designed to that will be concrete, bpotential fodegradation chemical re

NEI 142 8202 Editorial: Change “used” to 

“use.” 

Agree with comment Changed a

NEI 143 8228-8229 Delete this sentence. 

Requirements for water-tocement ratios and air content (mainly controlled by the use of air entraining admixtures), which are based on the severity of the anticipated exposure of concrete, are provided in ACI 349/318. The 

Agree with comment Deleted as 
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w/c ratio and air content are design requirements and not fabrication details. 

NEI 144 8301-8303 Samples normally taken in 

HAZ, same weld thickness and materials of construction, etc.: This area needs clarification. Testing is done per ASME Section III and Section IX. Weld thickness relation to the thickness of the design weld is governed by Section IX. Impact testing is required of the base metal (NX-2300 and the weld metal (NX-2400), but not the HAZ. Weld qualifications are performed using materials of the same class (P-number), but not necessarily the same material and grade as that used in construction. 

Agree with comment Replaced s

Metals havcubic crystaaustenitic sremain tougvery low tenot a conceAdded as sparagraph: Toughness Charpy impgoverned bas supporte

NEI 145 8319-8320 Specifying peak rod burnup is 

inconsistent with past practice, which has been to specify assembly average burnup. 

See response to NEI Comment 9 Changed a

Comment 9 

NEI 146 8358-8359 The text refers to “the following 

Part 72 regulations” yet no regulations are discussed in the text that follows. 

Agree with Comment. The 3rd par

8.4.17.1 wa

sentence in paragraph “The accepand review 

NEI 147 8453 Delete “and retrieval” since 

this is covered by fuel handling 

Agree with Comment Changed a

NEI 148 8567-8568 The text states that this review Agree with Comment Revised to 
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should be coordinated with the materials reviewer. The guidance in this section is specifically for the materials reviewer. Please clarify. 

with therma

NEI 149 8593-8595 Delete the last sentence of this 

paragraph. It is opinion, not review guidance. 

Agree with Comment Changed a

NEI 150 8636 Replace the word “dangerous” 

with “large” or “significant.” 

Agree with Comment The term “d

changed to 

NEI 151 8645-8656 Helium testing of the entire 

confinement boundary is not necessary. Confinement boundary welds are volumetrically tested in the fabrication shop and the entire vessel is pressure tested after loading. Both the inspections and testing are performed per the ASME Section III Code. Additional testing beyond what the ASME Code requires should not be necessary. Please revise. 

See response to NEI Comment 116. The ASME Code non destructive examinations and pressure test are performed for different reasons than the helium leak rate test which assures no significant radiological leakage. 

No Change 

NEI 152 8726 RG 1.183 should be RG 1.193. Agree with comment Changed aNEI 153 8914-8956 References to Part 71 

regulations do not appear appropriate in these lines. Please revise accordingly. 

Agree with comment Removed s

mentions PAlso, under damaged fuwas similar

NEI 154 8990-9013 Editorial: The numerals in the 

compound names should be subscripts to be consistent with the convention in other portions of the SRP. Please 

Agree with comment All chemica

paragraph written in sU4O9 but U
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Comment SRP Location Public Comment Resolution Cha

revise. 

NEI 155 9077 

9271 

Sections 8.7.3 and 8.8.3 should be removed and the references moved to the consolidated references in Appendix A to be consistent with the treatment of references in other chapters and to eliminate duplicate references (e.g. line 9089 and line 12923). 

Agree with comment Delete Sec

8.8.3. IntegAppendix A redundancy 

NEI 156 9090 Editorial: The reference 

incorrectly lists the upper temperature as 400. The correct value is 360 as listed in line 12923. 

Agree with comment Changed a

NEI 157 9231-9232 The limit could be interpreted 

as the limit in any one cycle is 65oC. It needs to explicitly 

state that the 65oC range can 

be exceeded but for less or equal to 10 cycles. 

This paragraph provides support for SRP Section 4.4.2, which discusses thermal cycling. 

No change 

NEI 158 9518 

9520 

Editorial: Sketches A and B should more appropriately be listed as Figures and the references to the sketches appropriately revised. 

Agree with comment Sketches A 

redesignateFigure 8-4, list of Figurto include t

NEI 159 9518 

9520 

Information was removed from the sketches when they were incorporated from ISG-18 Rev. 1 (e.g. identification of cover plate and vent and drain port cover plate). This information should be restored. 

Agree with comment Changed a

NEI 160 9737 Suggest changing “use and The SRP provides examples of the No Change 
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operation” to “function”. The cask vendor may not offer all of these specialized tools or require a particular tool to be used to accomplish a task. The user needs to understand the intended function for them to purchase the equipment needed to accomplish the task. 

specialized equipment and tools with enough detail for the staff to understand their use and operation (i.e. lifting yokes, transporter equipment, welding and cutting equipment and vacuum equipment). If their use and operation was changed to function, then the prior named examples would be sufficient since their name is self descriptive. The staff should review the description of how this specialized equipment is used and operated with the DSS, as stated in the SRP. 

NEI 161 9752 Delete “receipt inspection 

activities.” Receipt inspection is a separate QA function not related to the operations described in Chapter 9. 

Agree with comment Delete the 

inspection” listed 

NEI 162 7124-7125 9767-

9768 

Delete references to performing measurements to confirm assembly burnup values. Reactor records have repeatedly shown to be reliable for performing core reloads and to estimate boron concentration and rod position for reactor startup. They should be equally sufficient to validate assembly burnup for cask loading, a much lower risk activity. 

Current guidance on implementing burnup credit recommends a burnup verification measurement. This measurement is recommended to prevent misloading of assemblies that do not meet the loading criteria for burnup credit. The guidance indicates the level of agreement between the measurement and the record burnup values expected in order to confirm that the record is the correct record for the particular assembly being measured. Given the uncertainties in the 

Remove th7.5.5, “Burnreplace it wregarding tcredit, see of ISG-8. 
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measurement and record values, these uncertainties are to be accounted for in assigning the assembly’s burnup for loading into the cask. Based upon ACRS comments, as well as the expectation that the guidance on burnup credit will be revised in the near future, the guidance on burnup credit is being retained as a separate ISG (i.e., ISG-8) and the SRP is being modified to refer the reviewer to the current revision of this ISG. Revision to the ISG is expected to expand several aspects of the current guidance, including burnup verification. 

NEI 163 9847-9848 Delete the requirement to reevacuate and re-backfill. The 

necessary helium purity can be obtained with a single backfill of high enough purity. More generally, care should be taken in using the PNL document referenced because it is over 20 years old. Cask operations have changed in that time. For example, one current cask vendor dries the canister without the use of vacuum. We realize these are examples, but the reviewer should understand that the 

Agree with comment. Deleted the 

from the SRthen re-evabackfilled wfinal closurclarified to helium dryi
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reference document is out of date. 

NEI 166 9973-9974 Delete this item. Dose rates do 

not belong in TS and do not verify proper loading of the cask. 

Surface dose rate measurements are a parameter used to verify cask fabrication and operation. It is a measureable parameter during deployment of the cask onto the storage pad. Measurements may not detect all types of fabrication errors, but it provides a means for identifying potentially serious problems with the loaded contents and cask shielding system. 

The guidanwas revised measuremerates. 

NEI 167 10343 Editorial: Change “i.e.” to “e.g.” Agree with comment Changed aNEI 168 10345 Editorial: Delete close 

parentheses after “Program” and move the period inside the close quotation. 

Agree with comment Changed a

NEI 169 10366-10367 The “basis of tests deemed 

acceptable” should be from regulations or something more definitive and stable than prior staff acceptance. 

10CFR72.82, Inspections and Tests, Paragraph (d) states: “Each licensee shall perform, or permit the Commission to perform, such tests as the Commission deems appropriate or necessary for the administrator of the regulations in this part”. As such the regulations state the basis for performing the test as those deemed appropriate. The guidance also is specified to address unforeseen design proposals and considers the operational experience and precedent from previous licensing of licensing actions of storage casks.. 

Section 10The followipresented odeemed acin previous guidance is operational knowledge reviews. Acriteria may provides apexplanation justificationand criteria depending experience 
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the design 

NEI 170 10381-10382 Recurring trunnion load tests 

for transfer casks is not consistent with ANSI N14.6, which permits NDE to be performed periodically rather than load testing. 

The guidance for the load tests recognizes the lifting trunnion test provisions in accordance with ANSI N14.6. As such, periodical NDE, in lieu of annual load tests, is acceptable for the trunnion provided that other conditions as specified in ANSI N14.6 are also met 

Added the to the 1st pa

Section10.5Periodical Nannual load for the trunother condiANSI N14.6 

NEI 171 10418-10433 Please clarify the guidance 

pertaining to testing. Clarification should include ASME Code concurrence that fracture testing is not required for material with wall thicknesses of less than 5/8 inch. 

NUREG/CR-1815 "Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Up to Four Inches Thick" RG-7.11, ASME Code for Transport Packages and DOE guidance all establish fracture toughness testing for 3/16-inch and thicker material. Therefore, FT testing is required below 5/8-inch to down to 3/16-inch, unless other justification is provided. 

Changed ththe 4th para

10.5.1.1 to NUREG/CR“RecommeProtecting ABrittle FracShipping CInches Thicguidance cand thicknebrittle fractubasis of guNUREG/CR5.1.1, the Nmethods fomaterials.

NEI 172 10476-10479 Delete the sentence pertaining 

to inspection personnel qualifications. This is something governed by the QA program and outside the scope of a cask design review. At a minimum, delete “the current revision of.” The 

Partially agree with comment. The reference is appropriate but is changed from “current” to “appropriate” version. 

Deleted “cureplace it wrevision” in TC-1A 
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fabricator should not be forced to adopt the most recent revision of SNT-TC-1A to qualify personnel if a different code or older version of SNTTC-1A is acceptable within their QA program. If and when to adopt a later Code should be at their discretion. 

NEI 173 10513-10516 Why specify the particular 

NDE method if the Code does that? Suggest deleting this detail. Also, AWS should be offered as an acceptable weld code for non-confinement boundary welds. 

Agree with comment regarding AWS. Specific guidance on the Code requirements is provided to avoid misunderstanding and possible conflicting interpretations. In addition, the specific guidance assists reviewers in focusing on important elements of the NDE methods with respect to the associated review objectives. 

Added the last non-coparagraph “(LOW Prioconfinemenwelded, repin accordanStructural WSteel, D1.3Code – SheStructural WStainless Sstandards sthe licensin

NEI 174 10576-10577 Delete these lines. Dose rate 

measurements of every cask after SNF is loaded are of little value in determining whether the design criteria have been satisfied because the shielding analyses are extremely conservative. Users will perform appropriate dose rate measurements on the loaded casks as a part of their 

Partially agree with comment. The guidance is revised to indicate that dose measurements of loaded SNF, in lieu of an auxiliary source, may be used to verify shielding effectiveness with appropriate scanning of the shield and appropriate consideration of the actual source strength of the loaded contents. 

Revise Secinclude: Dose measSNF, in lieu source, mashielding efappropriate shield and program thactual sour
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Radiation Protection Program and ALARA procedures. 

loaded con

NEI 175 10588-10597 and 

10620-10629 

Duplicated paragraphs. Agree with comment Deleted 5th 

Section 10

NEI 176 10613 Editorial: “bench marked” 

should be “benchmarked” (one word). 

Agree with comment Changed a

NEI 177 10741 Clarify “periodic tests to verify 

shielding and thermal capabilities.” Such tests are usually not necessary for passively cooled systems beyond periodic checks of the air vents. Also, there are no credible age-related means to degrade shielding. Such tests should only be required if the particular cask design has unique features or active components requiring such tests. 

Aging and degradation of shielding materials may be a credible phenomenon. Degradations of components, such as cracks on concrete over-pack, corrosions of steel components, are examples that may impair their shielding capabilities. The applicant should otherwise justify that aging of materials related to the shielding, confinement, and thermal designs are not credible during the licensed period of the DSS. 

Section 10clarify that jrequired to confinemen

NEI 178 10955 Delete “including minors.” 

Minors are not part of the working staff at power plants subject to occupational exposure. 

Disagree with comment. In the event, that a minor is present in an occupational capacity at a licensee’s facility, that licensee is responsible for ensuring that the requirements of 10 CFR 20 are met. 

No Change 

NEI 179 10956 Delete “retrieval and”. The regulations (in 10 CFR 

72.236(h)) require that the spent fuel storage cask be compatible with both wet and dry loading and unloading facilities. It is reasonable to expect that fuel cannot be unloaded without first 

No Change 
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retrieving the storage cask from the storage location. The potential exists for there to be differences in the radiological conditions encountered in retrieving and unloading fuel from an ISFSI as compared to loading and emplacing the casks. The applicant should consider the possible differences in radiological conditions and provide dose estimates if any differences are expected to be significant. 

NEI 180 11003-11005 The value of applicants 

calculating and NRC approving dose versus distance from a hypothetical ISFSI is of questionable value in the application because of the arbitrary nature of: the number of casks, the arrangement of the casks on the ISFSI, the distance to the site boundary, and the cask contents. Licensees are required to perform a 72.104 dose analysis for their particular ISFSI by 72.212. 

The application should demonstrate that there is reasonable assurance that the requirements of 10 CFR 72.234(a) will be met by the proposed system. One of these requirements is presented in 10 CFR 72.236(d) which indicate that shielding and confinement features must be sufficient to meet the requirements of sections 72.104 and 72.106. The applicant must demonstrate the proposed system is capable of meeting these requirements. Past review experience has shown that the dose rate versus distance calculation for a standardized array has been beneficial in confirming these requirements. 

Replaced 2

Section 11following: The reviewthe applicarate versus evaluation hypotheticatheoretical consist of acasks (2x1analysis maof shielding array. The examine prand comparates from casks, and annual dosobserved foexisting ISF
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NEI 181 11007-11018 As only hypothetical array and 

single cask are evaluated, it is not clear when features would be required to show compliance with regulations and should be included in the conditions of cask use. Specific distance and shielding options and inclusion of such limitations in the CoC are not consistent with the 72.212 evaluation that a site would do to establish compliance with the requirements. 

As indicated in response to the previous comment, staff must be able to have reasonable assurance that shielding and confinement features for a proposed dry storage system are sufficient for users of the system to design ISFSIs that can meet the requirements of Sections 72.104 and 72.106. If the dose requirements of Section 72.104 can be met at the minimum distance specified in the regulations (100 meters, specified in Section 72.106(b)) for a single cask and the hypothetical array described in this section of the SRP, the staff considers there is reasonable assurance that the regulatory requirements will be met. If additional distance or shielding is needed to meet the dose limits beyond the controlled area boundary for either the single cask or the proposed hypothetical array, then staff needs some basis on which to make its determination. The applicant should provide a justification for how a general licensee could reasonably meet the requirements of Section 72.104. Including a shielding or distance requirement in the CoC conditions of use would only be needed if the applicant chose to use either or both of these as a basis for its SAR 

Moved the addition, thdetermine tthe normal could chanoff-normal end of the fSection 11Replaced t

and 4th para

11.5.3.1 wiparagraphsIt is importageneral ISFpermitted to between thcontrolled aengineered (supplemenas berms to real individThe SAR nsufficient ininformed chthe generaanalyses wthe minimudistance anadditional sprojected dexceeded tthe reviewethe applica
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evaluations and did not provide a SAR analysis without the added distance or shielding. Site-specific features, or extra distance or additional shielding that a general licensee chooses to evaluate and/or implement to further reduce doses outside the controlled area boundary are not included as limitations in the CoC, but are included in the site’s 72.212 evaluation. Analogous to the evaluations for 72.104, if evaluations for accidents show that a distance greater than 100 m or additional shielding is needed to meet the dose limits at or beyond the controlled area boundary, the preceding discussion would also apply for meeting the requirements of Section 72.106. Thus, in responding to this comment, staff also modified the guidance regarding accidents in Section 11.5.3.2 of the SRP to include similar guidance as provided for normal conditions in Section 11.5.3.1 of the SRP. 

justification licensee cothe require72.104. If thdistance cucask and hthe SAR wedistances gor assumed feature, thecontain a cthat effect. An example may be sim“The use of require mo100-meter the ISFSI aarea boundfeatures (i.ewalls), or bdose limits can be metengineered to ensure thof 10 CFR such featurconsidered [ITS] and mdetermine tcategory.” If an enginein the SAR 
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that feature to be part osuch, it shothe CoC. Inserted the at the end o

of Section For those sthese curvefor greater engineered berms, a cothat describsection of tneeded, wibeing 72.10instance.

NEI 182 11265-11266 Clarify this statement. Not all 

DSS monitoring equipment is ITS. It is only ITS if it meets the definition of ITS in the NUREG based on its design function. Suggest: “DSS monitoring equipment is classified in accordance with NUREG/CR-6407...” This also conflicts with lines 1678 and 5347. 

Agree with comment and changed to be consistent with Lines 5347 and 1678. Pressure monitoring systems are not ITS because they cannot withstand the design basis loadings, nor are they required to be procured in accordance with ITS practices. Their failure does not result in an unsafe condition, but their failure in combination with another failure (e.g. confinement seal) could result in an unsafe condition which makes it a Category B item under the guidelines of NUREG/CR-6407. 

Changed th“DSS moni(such as a system) are important to classified athe guidelin6407, ‘ClasTransportaDry Spent FComponenImportance 95/0551)’ sdesigned nthe same reconfinemen
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whose failuwith anothein an unsaf

NEI 183 11364-11368 What is the purpose of 

capitalizing this text? 

The statement is meant to caution the reviewer about terms and conditions of the CoC and technical specifications. The capitalization is removed from the text. 

Revised the follows (no If a reviewedesign featanalytical aoperating alimiting conprogram or important achanged wapproval, thfurther evaconsidered condition ospecificatioshould furthguidance in establishing technical spCoC. Only conditions the attachespecificatioare legally reviewer deimportant thchanged wapproval, thbe included terms, condspecificatio
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NEI 184 11440-11452 Most of the text about the 

Code in this paragraph is of limited value. Suggest replacing this with simpler guidance that states the applicant should state the applicable design codes, sections, subsections, as appropriate, and any alternatives to the code being implemented. 

The referenced paragraph presents the current and historical basis for the use of the ASME Code for DSS as guidance for the staff. 

No Change 

NEI 185 11460 Editorial: Add “s” to the end of 

“specification.” 

Agree with comment Changed a

NEI 186 11588 Editorial: Change “12” to “13.” Agree with comment Changed aNEI 187 12723-12724 ISG-15 should not be listed in 

the reference section since it has been incorporated into this document. Other ISGs are not listed in the reference section. 

Agree with comment Changed a

NEI 188 13025 Editorial: Change “to” to “10.” Agree with comment Changed aNEI 189 13158 Editorial: Insert a close 

parenthesis at the end of this line. 

Agree with comment Changed a

 NEI 190 13237 Editorial: Change ‘uncorrectd” 

to “uncorrected.” 

Agree with comment Changed a

NEI 191 13475 Editorial: Change “austentic” to 

“austenitic.” 

Agree with comment Changed a

NEI 192 13475 With regard to ISG 12, the 

status block states that a new revision is pending. This is inappropriate information for the SRP. In addition, a pending revision to this ISG has not been announced by NRC, yet draft revisions to 

Agree with comment Deleted “ne
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ISG-2 and ISG-23 have been issued by NRC and are not noted in this appendix. 

NAC General The Draft NUREG-1536 does 

not appear to reflect NRC’s position on risk based regulations. It appears to be too prescriptive in areas that have little to no impact on safety Reconsider detailed prescription of requirements that are covered by other regulations, measurements and controls, e.g., shielding design, related computer verification, measurements required during loading operations, measurements on loaded casks for site operations to manage site boundary dose. Technical Specification material should be limited to system operational limits that the licensee must meet and not repeat regulatory requirements or include material property and test requirements addressed by Quality Assurance requirements. Technical specifications should not be used as a control on the licensee use of 

NRC staff is not revising the Part 72 regulatory requirements as part of the update to this SRP. Some Part 72 regulations are prescriptive and others are performance based. In addition, the NRC does not endorse a risk-based approach, but rather a risk-informed approach as delineated in SECY-98-144. The SRP is being revised to risk-inform, or risk prioritize the review guidance used to verify that the established Part 72 regulations are met. The areas of review in the SRP have been prioritized considering the potential relative risk impact of not meeting the requirements. It provides guidance that reduces the intensity of the review for low risk areas. It is in this sense of focusing staff attention on areas important to safety that the SRP is riskinformed. In fact, this is the NRC definition of risk-informed. Also see response to NEI 18 regarding Technical Specifications. 

To clarify thstaff is usinreview procthis SRP anconfusion wquantitative text has begenerally sfor “risking referring to procedureshas been s
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72.48 revisions. 

NAC 1914-1917 “Nevertheless, for assessment 

purposes and to demonstrate ... the DSS should be evaluated for effects of a confinement boundary failure.” This is not duplicated in confinement SRP discussion. Evaluation of the effect of a confinement boundary failure is not a standard evaluation set for current licensed systems (ISG-5). Nonmechanistic failure should not be a system analysis requirement. This imposed analysis is beyond regulation requirements. 

See NEI Comment 28 See NEI Co

NAC 3106, 12537 Imposes excessive 

conservatism for seismic evaluations. RG 1.60 should be replaced by NUREG/CR-6728 and also NUREG/CR-6865. 

See NEI Comment 46. See NEI Co

NAC 3139-3140 “Confinement casks” is poor 

terminology. It should read: “for the confinement boundary of the cask.” 

Agree with comment Changed a

NAC 3153 In the previous paragraph, 

Subsection NB is used to define stress qualification for the confinement boundary, which is a pressure retaining 

See NEI Comment 48 See NEI Co
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boundary. In this paragraph, it does not clearly state that the basket is a nonpressureretaining boundary and that the applicant should use Subsection NG. Need to state that Subsection NG is acceptable or else the reader is left to believe that Subsection NB applies to nonpressure boundary baskets. It should confirm that Appendix F is applicable for use with Subsection NG. 

NAC 3168 Includes excessive 

conservatism that is not consistent physical testing. It should state that Subsection NB and Subsection NG permit the use of Appendix F, which does permit the use of inelastic properties for components that serve as the pressure boundary or also non-pressure boundary applications, such as baskets. It should also state that strain base criteria can be employed for energy limited accident conditions, provided the applicant provides such basis for its use.“ 

The strain-based criteria are not recognized by the ASME Code or other applicable standards. The NRC staff may consider use of other acceptance criteria on a case-by-case basis. 

No change 

NAC 3171 In many applications for drop The SRP does not preclude use of No change 
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conditions, it should be acceptable to strain rate sensitive properties. Appendix F permits its use. Need to include “strain rate properties, which need the appropriate references.” 

strain-rate-sensitive material properties for design analysis of cask drop conditions 

NAC 4302 Annotation of input files. It is 

important to be able to use the applicant’s files. It is not necessary to understand all aspects of the input files. Some of these files come from Journal files or Log files which are generated by the program. It is not feasible to add comments to these files. Open-ended statements such as adding “annotation” lead to overstatement by the reviewer for the need of such documentation. 

See NEI Comment 57 See NEI Co

NAC 4313 Annotation of the load steps. 

This would lead to excessive documentation in the computer solutions. 4311-4315 should be removed. It is the responsibility of the applicant’s QA program to ensure that the analyses are performed correctly. 

See NEI Comment 58 See NEI Co

NAC 4332 Sensitivity study on mesh type. 

Lack of clarity. “Mesh type” should be removed. It is not 

See NEI Comment 59 See NEI Co
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clear. 

NAC 4335-4336 Mesh study. Not required 

when stress linearization is being used for primary loading. Such detailed studies should be restricted to fatigue evaluations at stress discontinuities. Remove these lines. Too subjective, allowing the reviewer to specify detailed mesh studies for any part of the model he so desires. 

See NEI Comment See NEI Co

NAC 4349 Including plots of the results. 

Generates extra data to be included in the SAR, while it is not needed. Remove “plots” from line 4349. 

See NEI Comment 61 See NEI Co

NAC 4680 Exclusion of natural 

convection internal to the canister. Too restrictive for convection designs. It states: “...should be limited to...the external surface...” This is an unacceptable statement that will be taken by the reviewer that internal convection cannot be used without some excessive burden of proof provided by the applicant. Remove line 4680. There is sufficient test data to confirm that convection internal to the canister is acceptable. 

See NEI Comment 69 See NEI Co

NAC 4687/JS Convection. What does 

“robust” mean? This allows the 

See NEI Comment 71 See NEI Co
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reviewer to apply his personal definition of “robustness” to the applicant’s analyses. Remove “robust” from Line 4687. 

NAC 5185 Confinement Monitoring 

Capability. Welded closure seal. “However, the lack of a closure monitoring system has typically been coupled with a periodic surveillance program that would enable the licensee to take timely and appropriate corrective actions ...” Dry cask storage systems have been approved without a closure weld seal monitoring system, as within the storage cask, surveillance of the closure weld is not feasible. Temperature monitoring and/or visual surveillance of the air cooling vents is a standard part of concrete cask (welded canister) licensing. 

See NEI Comment 73 See NEI C
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NAC 5426 Table 5-2, Release Fractions. 

“...should not be used for spent fuel described as damaged.” Based on NUREG/CR-6497, damaged fuel would not have a driving force to release fines from the matrix. What is the postulated issue here? Is there data available to NRC that indicates a safety concern? Provide additional guidance and describe what factors are suggested for damaged fuel. 

There is lack of data regarding the release fractions from damaged fuel, which makes this a safety issue. The data available does not apply to damaged fuel but rather to a single breach of one fuel rod. Compounding the issue is that many of the storage canisters are pressurized with helium to aid in the heat removed of larger thermal payloads. Without compelling factual information and data regarding the release fractions associated with damaged fuel, the staff does not feel there is not adequate evidence to generically make assumptions regarding release fractions associated with potential types of damaged fuel. Therefore, a leaktight confinement boundary is the recommended accepted practice to ensure radiological safety for damaged fuel, without additional data and analyses from the applicant. 

See NEI Co

NAC 5799-5801 Radiation Source Definition. 

"radionuclide content, and estimated radiation source strength in Becquerels, .... should be described...." New requirement. Provide clarification as to what the basis of this request is, as radiation source strength in Ci 

See NEI Comment 77. See NEI Co
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or Bq is not clearly related to gamma/neutron source strength (e.g. beta emitters) 

NAC 5809-5810 Radiation Source Definition 

(Gamma Sources) "characteristics for each gamma-ray source type should be provided, including isotopic composition, and photon yields” Is a tabulation of spent fuel isotopics requested here? If so, to what purpose? Typically, inputs into depletion analysis are provided, but not isotopics of depleted materials. Clarify requirement if a tabulation of spent fuel isotopics is requested and describe purpose 

See NEI Comment 78 See NEI Co

NAC 5813-5814 Radiation Source Definition 

(Gamma Sources) Within gamma source description, "describe the extent to which radioactivity may be induced by interactions involving neutrons originating in the stored materials" If this implies n-gamma reactions, then the current SRP version is clearer If activation is to be considered for decommissioning, that should be clarified. 

See NEI Comment 79 See NEI Co
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NAC 5868-5870 Shielding Analyses (Computer 

Codes) "The applicant should defend any simplifications and assumptions by showing that the approach used will result in conservative (bounding) estimates.” Clarify if results need to be bounding or "provide reasonable assurance" as stated in Section 6.4, Line 5723: "Reasonable assurance that the proposed design fulfills the acceptance criteria " 

See NEI Comment 80. See NEI Co

NAC 5873-5874 Shielding Analyses (Computer 

Codes) "...SAR should numerically specify source term uncertainties for high burnup fuels" in combination with "...validation data is relatively limited for burnup above 45 GWd/MTU.“ High burnup fuel is licensed and in storage. No indication that substantial dose effects occurred. If limited data is available it leaves an open ended question as to how to specify uncertainties. Conservative assumption and desired design margins are not defined, leaving it up to each reviewer when, and how much, in uncertainties to 

See NEI Comment 81. See NEI Co
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apply. Provide correlation why maximum fuel assembly heat load is related to uncertainties. Low heat capacity/minimal shield system may be affected by low fuel assembly heat load and vice versa 

NAC 6003-6004 Radiation Source Definition 

(Initial Enrichment) "Applicant and the staff should not attempt to establish specific source terms as operating control and limits for cask use." If that is the case, why does the SRP focus in the Section 6.4.2 on curie content and isotopic description of the spent fuel? For Cobalt-60 dominated hardware sources, a source term may be more appropriate than other limits (e.g., mass, exposure, cool time). 

See NEI Comment 84 See NEI Co

NAC 6149-6150 Shielding Model Specification 

(Configuration of the Shielding and Source) "...homogenization should not be used in neutron dose calculation when significant neutron multiplication can result from moderated neutrons..." While not changed from current SRP statement, it 

See NEI Comment 84 See NEI Co
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should be noted that standard practice is to homogenize the rod lattice in shielding calculations (not necessarily homogenizing basket structure into the fuel region). Provide additional guidance and/or justification why the standard practice of homogenizing the rod lattice in shielding calculations should not be used. 

NAC 6221-6222 Shielding Analyses 

(Computer Codes) "The reviewer should be aware that often adjoint calculations are performed by the applicant ... importance functions..." Review staff should recognize that importance functions may also be produced with Monte Carlo, point-kernel and transport codes. Include importance functions produced with Monte Carlo, point-kernel and transport codes 

See NEI Comment 88 See NEI Co

NAC 6246-6248 Shielding Analyses 

(Computer Codes) “The applicant should use the latest released computer code version that is valid for the particular computational platform used to perform the analysis." This item in particular has been discussed 

See NEI Comment 89 See NEI Co
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with NRC staff as a significant issue. Licensed code for same type of application should not require code version change unless safety issue has been identified. Continual use of different code version within an application is difficult to reconcile and potentially leads to unnecessary confusion. Typical new release code versions tend to contain a certain amount of bugs that get resolved through user feedback to code originator. Could be interpreted that a newer code provides more "accurate“ result; but as previous version was found to be acceptable for system approval, there should be no requirement for change. The goal per draft SRP Section 6.4 is to provide reasonable assurance that system will meet limits. 

NAC 6302-6309 Shielding Analyses 

(Computer Codes) "by verifying that the following information has been provided in the SAR ... The computer code solutions to a series of 

See NEI Comment 90 See NEI Co
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test problems ..." Draft SRP does not contain the previous SRP statement "that these solutions may be referenced, and need not be submitted in the SAR". This change would add a substantial amount of information to the SAR without any safety benefit, as the referenced documents, per current SRP, should be public information and/or have been previously submitted to NRC. Adopt current SRP verbiage and add: “These solutions may be referenced but need not be submitted in the SAR.” 

NAC 7697 Methodology to Evaluation 

Flaw Size For stainless steel casks and welding, this is too limiting. NAC uses the J-integral method to evaluate flaw size which limits the size of a single weld pass. In order to be consistent with 7682, it should explicitly state that the applicant can use J integral methodology based incorporating plasticity for ductile weld materials such as stainless steel. 

See NEI Comment 121 See NEI Co
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NAC 9131-9232 Fuel Temperature Range 

Limits This could be interpreted as the limit in any one cycle of fuel temperature is limited to 65°C. It needs to explicitly state that the 65°C range can be exceeded, but for less or equal to 10 cycles. 

See NEI Comment 157 See NEI Co

NAC 10418-10433 Charpy Test Requirements. 

Use of carbon steel less than 5/8 inch thickness. NRC’s position/guidance should be stated. Clarification should include ASME Code concurrence that fracture testing is not required for material with wall thicknesses of less than 5/8 inch 

See NEI Comment 171 See NEI Co

NAC 11007 Exposures at or Beyond the 

Controlled Area Boundary 

(Normal Conditions) Focus added on "additional engineering features and distance from array." As only hypothetical array and single cask are evaluated, it is not clear when features would be required to show compliance with regulations and should be included in the conditions of cask use. Specific distance and shielding options and inclusion of such 

See NEI Comment 181 .See NEI C
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limitations in the CoC do not seem to be consistent with the 72.212 evaluation that a site would do to establish compliance with the requirements. Further guidance is required 

NRC 

Clarification 

1445, 1507, 1659, 1682, 1701, 2501, 4533, 4590, 5094, 5249, 

6066, 6648, 

9993, 11242, 

Corrected the typo for the word “Principal” 

Correct typo. Change SR

NRC 

Clarification 

5765 Deleted the word “rate”. 

Deleted the words “for occupational exposure and”. 

Revise the statement to make it more accurate. 

Changed S

NRC 

Clarification 

5793 Deleted the words “placed” Revise the statement to make it 

more accurate. 

Changed S

NRC 

Clarification 

5873 Changed the verb “is” to “are” Correct typo.. Changed S

NRC 

Clarification 

5880 Deleted the word “each” Correct grammar error Changed S

NRC 

Clarification 

5908 Deleted the word “operations” Delete redundant word Changed S

NRC 

Clarification 

5915 Replaced the word “functions” 

with “operations” 

Revise the statement to make the meaning clearer 

Changed S

NRC 

Clarification 

5960 Changed the words “5” to “6” Correct typo. Changed S

NRC 

Clarification 

5977 Corrected the typo for the 

word “Principal” 

Correct typo. Changed S

NRC 

Clarification 

5999 Corrected the typo for the 

word “Principal” 

Correct typo. Changed S

NRC 

Clarification 

6018 Changed “fall” to “falls” Correct typo Changed S
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Comment SRP Location Public Comment Resolution Cha

NRC 

Clarification 

6103 Added verb “are” between 

“there” and “specific” 

Correct typo. Changed S

NRC 

Clarification 

6230 Changed “ORNL” to “EPRI” Correct typo. Changed S

NRC 

Clarification 

6373-6379 Modified line 6373 to add the 

following words between “to use” and “additional ...”: “distance or” Modified line 6374 to add the following words between “berms,” and “to mitigate”: “or both,” Modify line 6375 to replace the words “to show compliance with the regulations” with “evaluations” Modify line 6376 to replace the words “cask conditions for use” with “system and described in the CoC” 

Revised the statement to make them more accurate and consistent with staff’s response to NEI’s Comment 181 

Changed S

NRC 

Clarification 

6405 Changed “1m (3.3ft)” to “100m 

(328 ft)” 

Corrected typo. Changed S

NRC 

Clarification 

6408 Delete the word 

“phenomenon” 

Revised the statement to make it consistent with 10 CFR 72.92. 

See NEI CoSRP as ind

NRC 

Clarification 

8643 Changed priority of Section 

8.4.20 to Medium from Low 

Change was result of ongoing industry practice of omitting leak testing at fabrication facility 

Added “(MEafter Sectio

NRC 

Clarification 

10976-10979 This sentence mis-states what 

is required by the regulation. 

Reword the sentence so that it correctly reflects the regulatory requirement. 

Replaced tthe first par11.5.3 with As required 72.236(d), 
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demonstratand confinecask are surequiremenindividuals and for DBA CFR 72.10demonstratapplication specific evageneral ISF

NRC 

Clarification 

10980 Added the word “boundary” 

after “controlled area” 

Revised the statement to make the meaning clearer

Changed S

NRC 

Clarification 

6863,6864,6993, 

6994,7005, 

7098,7100 

Editorial corrections are needed to make the citation for references compatible with the format of the Consolidated References in Appendix A. 

Comment implemented. See NEI Comment 86.Correct typo. 

Changes to citations weindicated to the format 

NRC 

clarification 

 The meaning of the first 

statement in Section 6.1 is unclear regarding what is evaluated for the review. 

The statement was clarified to indicate that the evaluation is of the ability of the shielding features to provide adequate shielding. 

Modified thSection 6.1 The shieldithe ability oshielding dprovide adeagainst diredry storage contents. 

NRC 

clarification 

 The statement in the third 

paragraph of Section 6.2 refers to dose rate limits; this is not correct. The regulatory limits are dose limits 

The statement was corrected to refer to dose limits. 

Removed tthe sentencparagraph bottom pagsentence wdose limits 
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area bound

NRC 

clarification 

 The shielding chapter 

guidance covers all of 72.104, not just 72.104(a). Table 6-1 should indicate this. 

Table 6-1 was corrected to refer to 72.104. 

Deleted “(aTable 6-1. 

NRC 

clarification 

 The terms “general licensee” 

and “DSS user” refer to the same entity; so, using them together, as is done in the first paragraph of Section 6.4, is redundant. 

The sixth sentence of the paragraph was modified to delete the redundant reference. 

Deleted “Dsixth senteparagraph 

NRC 

clarification 

 The guidance described as 

part of this item 1. in Section 6.4 (second paragraph) includes all of 72.104 and not just 72.104(a). 

The statement was corrected to refer to 72.104. 

Deleted “72sentence o6.4, second replaced it 

NRC 

clarification 

 Part 72 establishes dose 

requirements for ISFSI operations (first paragraph in Section 6.5.1.1, second sentence). 

The sentence was clarified to indicate that ISFSI operations are included in the dose requirements. 

Modified thof the first p6.5.1.1 to rWhile 10 Cestablishes for the ISFSit does not dose rate liindividual c

NRC 

clarification 

 The box for Chapter 9 in 

Figure 6-1 shows an item for pre-shipment surveys. This language seems to be inappropriate since there is no shipment of the cask under Part 72. 

The figure was corrected to remove references to pre-shipment surveys. Surveys are still a part of operations; so, the surveys are linked with the item for radiation levels. 

Deleted “PrSurveys” frbox in Figuthe first bul“Radiation 

NRC 

clarification 

 The limits in 10 CFR 72.104 

are for doses and not dose 

The reference to the regulation was corrected to state dose limits. The 

Deleted the the third se
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Comment SRP Location Public Comment Resolution Cha

rates. Also, applicants may also use dose v. distance curves for evaluations for this requirement instead of dose rate v. distance curves. (See fourth paragraph of Section 6.5.4.3.) 

text is also clarified to indicate that dose v. distance or dose rate v. distance curves may be used for evaluations for this requirement. 

paragraph Add the wofifth sentenparagraph begins with should inclurate versus 

NRC 

clarification 

 Section 72.104(c) also applies 

to operations. Section 72.104 applies to all operations; there is nothing in the regulations to indicate that Section 72.104 doesn’t apply to cask unloading. (See Table 9-1.) 

Table 9-1 was corrected to include reference to 72.104(c) and show that 72.104(b) and (c) apply to cask unloading. 

Included “7of applicabTable 9-1 athe table enand (c) and 

NRC 

clarification 

 The text in Section 10.5.1.4 

should be clear that the testing is for materials used for neutron shielding; guidance for testing neutron absorbers for criticality control is described in the following subsection of the SRP 

The text in the second paragraph of the section was clarified to focus on neutron shielding materials. 

Deleted the and replacein the first ssecond par10.5.1.4. 

NRC 

clarification 

 The guidance in Sections 

10.5.1.6 through 10.6 applies to the DSS. The text should be clear and use this terminology. 

The text of these sections was corrected to refer to the DSS. 

Replaced tthe word “Dbeginning oin Section 1the word “cappeared athe sentencword “casktitle of Sectthe first sen10.5.1.7. Replaced tcasks” with 
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Comment SRP Location Public Comment Resolution Cha

sentence in Replaced tcask” with “sentence oSection 10word “caskfirst senten10.5.2.3. R“cask” with F10.3, F10second sensummary, p10.6. 

NRC 

clarification 

 The second sentence of the 

summary paragraph in Section 10.6 is missing the object of what is being stored (i.e., the words “spent fuel” are missing). 

The sentence was corrected to include the words “spent fuel”. In conjunction with the previous clarification comment, the word “cask” was replaced with “DSS.” 

Corrected tof the summSection 10DSS will alspent fuel tlicensed or 

NRC 

clarification 

 Table 11-2 is missing 

72.126(d), which is also applicable to radiation protection. 72.126(a) and (d) apply to all areas of review listed in the table. Also, there is no 72.1064(b); it should be 72.106(b) and ALARA does not apply to 72.106(b). 

Table 11-2 was corrected to include the stated items. 

Include “72bullets to sof 72.126(aoccupationexposures controlled a“72.1064(bread “72.10reference to removed.. 

NRC 

clarification 

 The last sentence in Section 

11.4 lists reviews that should be coordinated with the Chapter 11 review; “the chapters” should read “these chapters”. 

This editorial correction was implemented. 

Edited the Section 11“...thereforthese chapcoordinated
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Comment SRP Location Public Comment Resolution Cha

NRC 

clarification 

 Section 11.4.2 should be clear 

on what points may be included in the representative points in the restricted areas discussed in the first paragraph. Also there may be only one restricted area at a general licensee’s site. 

Examples were provided to illustrate some representative points in the restricted areas. The term “areas” was modified to area(s) to capture the idea that sites may or may not have more than one restricted area. 

Changed th“area(s)” anand near thstorage oveafter the woarea(s)”. 

NRC 

clarification 

 The Chapter 12 box in Figure 

11-1 shows “Estimated Dose Rates” twice. 

The second bullet for estimated dose rates was deleted. 

Deleted the Dose Rates12 box in F

NRC 

clarification 

 The first paragraph in Section 

11.5.3.1 incorrectly refers to dose rates in 10 CFR 72.104. Also applicants may use dose versus distance or dose rate versus distance curves for these evaluations (see first two paragraphs of Section 11.5.3.1). 

The reference to regulatory dose rates was corrected to read as regulatory doses, and the text was clarified to account for use of dose versus distance curves in the applicant’s evaluations. 

Changed “d“doses” in tthe first par11.5.3.1. Changed ththe first par11.5.3.1 to dose rate vChanged ththe second Section 11“...applicandose rate v

NRC 

clarification 

 The full title of the referenced 

document should be given in Section 12.5.2. 

The text was corrected to give the full document title in the second paragraph of this section. 

Modified ththe second Section 12“...NUREG“ClassificatPackaging Storage SyAccording tSafety.””. 
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Comment SRP Location Public Comment Resolution Cha

NRC 

clarification 

 SAR sections are the 

applicant’s evaluation and not review sections. (See the second paragraph of Section 13.1) 

This editorial correction was made. Corrected t

the second Section 13“...justified evaluation ...” 

NRC 

clarification 

 In Section 13.5 (see the third 

bullet in the fourth paragraph), the term maximum burnup should be just stated as burnup. Appropriate specifications may need to include either a maximum burnup or a minimum burnup or both depending upon the proposed contents and system design. Also, maximum neutron and gamma source terms are not an adequate specification to define proposed spent fuel contents and aren’t being accepted as such. 

The text of referenced bullet was corrected as described in the comment. 

Deleted the that occurs “burnup” anmaximum nsource termbullet (in thin Section 1

NRC 

clarification 

 Two bullets without text 

appear in the bulleted list in Section 13.5. 

These two bullets were removed. Deleted the 

text in the l(fourth para