Standard Review Plan

for Renewal of Specific 

Licenses and Certificates of 

Compliance for Dry Storage 

of Spent Nuclear Fuel

Final Report

Office of Nuclear Material Safety and Safeguards

NUREG-1927

Revision 1

Standard Review Plan

for Renewal of Specific 

Licenses and Certificates of 

Compliance for Dry Storage 

of Spent Nuclear Fuel

Final Report

Manuscript Completed: May 2016

Date Published: June 2016

Office of Nuclear Material Safety and Safeguards

NUREG-1927

Revision 1

iii

ABSTRACT

This Standard Review Plan is intended for use by the U.S. Nuclear Regulatory Commission 

(NRC) reviewer. It provides guidance for the safety review of renewal applications for specific 

licenses of independent spent fuel storage installations and certificates of compliance (CoCs) of

dry storage systems, as codified in Title 10 of the Code of Federal Regulations (10 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.”

This Standard Review Plan provides guidance for the review of general information, scoping 

evaluation information, and aging management information, included in a renewal application. 

The guidance provides information on review of time-limited aging analyses and aging 

management programs (AMPs), including learning AMPs that consider and respond to 

operating experience. The guidance provides example AMPs for welded stainless steel 

canisters, reinforced concrete structures, and a high burnup fuel monitoring and assessment 

program. It also provides guidance on considerations for CoC renewals and the general license 

framework, including guidance on general licensees’ implementation of AMPs.

The NRC expects to periodically revise and update this Standard Review Plan to clarify the 

content, correct errors, and include new information, knowledge, and experience regarding 

aging management considerations. Comments, suggestions for improvement, and notices 

of errors or omissions should be sent to the Director, Division of Spent Fuel Management, 

U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

Paperwork Reduction Act Statement

This NUREG contains information collection requirements that are subject to the Paperwork 

Reduction Act of 1995 (44 U.S.C. 3501 et seq.). The Office of Management and Budget (OMB) 

approved these information collections under OMB control number 3150-0132.

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for 

information or an information collection requirement unless the requesting document displays a 

currently valid OMB control number.

v

CONTENTS

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

LIST OF FIGURES ...................................................................................................................vii

LIST OF TABLES.....................................................................................................................vii

ABBREVIATIONS .....................................................................................................................ix

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

1. GENERAL INFORMATION REVIEW.................................................................................... 7

1.1 Review Objective............................................................................................................ 7

1.2 Areas of Review ............................................................................................................. 7

1.3 Regulatory Requirements............................................................................................... 7

1.4 Review Guidance ........................................................................................................... 8

1.4.1 General Information .............................................................................................. 8

1.4.2 Financial Information ............................................................................................ 8

1.4.3 Environmental Report ........................................................................................... 9

1.4.4 Application Content............................................................................................... 9

1.4.5 Timely Renewal ...................................................................................................11

1.4.6 Amendment Applications Submitted during the Renewal Review or after the 

Renewal Is Issued ...............................................................................................11

1.4.7 Terms, Conditions, and Specifications for Specific Licenses and CoCs in the 

Period of Extended Operation..............................................................................12

1.5 Evaluation Findings .......................................................................................................14

2. SCOPING EVALUATION ....................................................................................................15

2.1 Review Objective...........................................................................................................15

2.2 Areas of Review ............................................................................................................15

2.3 Regulatory Requirements..............................................................................................15

2.4 Review Guidance ..........................................................................................................16

2.4.1 Scoping Process..................................................................................................16

2.4.2 Structures, Systems, and Components within the Scope of Renewal ..................18

2.4.2.1 Scoping of Fuel Assemblies..................................................................19

2.4.2.2 Scoping of Structures, Systems, and Components, Depending on 

Individual Design Bases........................................................................19

2.4.3 Structures, Systems, and Components Not within the Scope of Renewal............19

2.5 Evaluation Findings .......................................................................................................20

3. AGING MANAGEMENT REVIEW .......................................................................................21

3.1 Review Objective...........................................................................................................21

3.2 Areas of Review ............................................................................................................21

3.3 Regulatory Requirements..............................................................................................23

3.4 Materials, Service Environments, Aging Mechanisms and Effects, and Aging 

Management Activities ..................................................................................................25

3.4.1 Review Guidance.................................................................................................25

3.4.1.1 Identification of Materials and Environments .........................................25

3.4.1.2 Identification of Aging Mechanisms and Effects ....................................26

3.4.1.3 Aging Management Activities ................................................................28

3.4.1.4 Aging Management Review for Fuel Assemblies ..................................28

3.4.2 Evaluation Findings .............................................................................................29
vi

3.5 Time-Limited Aging Analyses ........................................................................................30

3.5.1 Review Guidance.................................................................................................31

3.5.2 Evaluation Findings .............................................................................................32

3.6 Aging Management Programs .......................................................................................32

3.6.1 Review Guidance.................................................................................................33

3.6.1.1 Scope of Program.................................................................................33

3.6.1.2 Preventive Actions ................................................................................33

3.6.1.3 Parameters Monitored or Inspected ......................................................34

3.6.1.4 Detection of Aging Effects.....................................................................34

3.6.1.5 Monitoring and Trending .......................................................................35

3.6.1.6 Acceptance Criteria...............................................................................36

3.6.1.7 Corrective Actions.................................................................................36

3.6.1.8 Confirmation Process............................................................................37

3.6.1.9 Administrative Controls .........................................................................37

3.6.1.10 Operating Experience ...........................................................................38

3.6.2 Commencement of AMP(s) for CoC Renewals ....................................................40

3.6.3 Implementation of AMP(s) ...................................................................................40

3.6.4 Evaluation Findings .............................................................................................40

4. CONSOLIDATED REFERENCES .......................................................................................43

5. GLOSSARY.........................................................................................................................47

APPENDIX A NON-QUANTIFIABLE TERMS .......................................................................A-1

APPENDIX B EXAMPLES OF AGING MANAGEMENT PROGRAMS..................................B-1

APPENDIX C [RESERVED] ..................................................................................................C-1

APPENDIX D SUPPLEMENTAL GUIDANCE FOR THE USE OF A DEMONSTRATION 

PROGRAM AS A SURVEILLANCE TOOL FOR CONFIRMATION OF 

INTEGRITY OF HIGH BURNUP FUEL DURING THE PERIOD OF 

EXTENDED OPERATION...............................................................................D-1

APPENDIX E CONSIDERATIONS FOR RENEWALS OF CERTIFICATES 

OF COMPLIANCE ..........................................................................................E-1

APPENDIX F STORAGE TERMS .........................................................................................F-1
vii

LIST OF FIGURES

Figure A Specific-license and CoC renewal process ............................................................. 3

Figure 2-1. Flowchart of scoping evaluation process................................................................17

Figure 3-1. Flowchart of AMR process .....................................................................................22

Figure F-1. Flowchart for calculating storage terms................................................................ F-2

LIST OF TABLES

Table 1-1. Relationship of Regulations and General Information Review ................................. 7

Table 2-1. Relationship of Regulations and Scoping Review ..................................................16

Table 3-1. Relationship of Regulations and Aging Management Reviews...............................24

Table A-1. Screening Criteria for Non-Quantifiable Terms.....................................................A-2

Table B-1. Example AMP for Localized Corrosion and Stress Corrosion Cracking of 

Welded Stainless Steel Dry Storage Canisters.....................................................B-3

Table B-2. Example AMP for Reinforced Concrete Structures.............................................B-16

Table B-3. Example of a High Burnup Fuel Monitoring and Assessment Program...............B-24

Table B-4. Formal Evaluations of Aggregate Information on HBU Fuel Performance ..........B-29

ix

ABBREVIATIONS

ACI American Concrete Institute

AMA aging management activity

AMP aging management program

AMR aging management review

ANSI American National Standard Institute

ASME American Society of Mechanical Engineers

°C degrees Celsius

CAP Corrective Action Program

CFR Code of Federal Regulations

CISCC chloride-induced stress corrosion cracking

CLB current licensing basis

CoC certificate of compliance

CRIEPI Central Research Institute of Electric Power Industry

DOE U.S. Department of Energy

DSS dry storage system

EPRI Electric Power Research Institute

°F degrees Fahrenheit

FSAR final safety analysis report

GTCC greater-than-Class-C

GWd/MTU gigawatt days per metric ton uranium

HBU high burnup

HDRP HBU Dry Storage Cask Research and Development Project

ICSF interim consolidated storage facility

ISFSI independent spent fuel storage installation

ISG Interim Staff Guidance

ITS important to safety

kJ kilojoule

LWR light water reactor

m meters

mol mole

NDE nondestructive examination

NEI Nuclear Energy Institute

NMSS Office of Nuclear Material Safety and Safeguards

NRC U.S. Nuclear Regulatory Commission

OMB Office of Management and Budget
x

PM NRC Project Manager

ppm parts per million

QA quality assurance

s second

SAR safety analysis report

SER safety evaluation report

SRP Standard Review Plan

SSC structure, system, and component

TEPCO Tokyo Electric Power Company

TLAA time-limited aging analysis

TS technical specification

UFSAR updated final safety analysis report
1

INTRODUCTION

This Standard Review Plan (SRP) is intended to provide guidance to the U.S. Nuclear 

Regulatory Commission (NRC) staff for the safety review of renewal applications for specific 

licenses of independent spent fuel storage installations (ISFSIs) and certificates of compliance 

(CoCs) of dry storage systems (DSSs), as codified in Title 10 of the Code of Federal 

Regulations (10 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.”

To renew a specific license, an applicant (i.e., the licensee) must submit a license renewal 

application at least 2 years before the expiration of the license, in accordance with the 

requirements of 10 CFR 72.42(b). To renew a CoC, an applicant (i.e., CoC holder, user, or 

user’s representative) must submit a renewal application at least 30 days before the expiration 

of the associated CoC in accordance with the requirements of 10 CFR 72.240(b). The NRC 

may renew a specific license or a CoC for a term not to exceed 40 years, in accordance with 

10 CFR 72.42(a), or 10 CFR 72.240(a), respectively.

The NRC-approved DSSs listed in 10 CFR 72.214, “List of Approved Spent Fuel Storage 

Casks,” may be used by any 10 CFR Part 72 general licensee in accordance with 

10 CFR 72.212, “Conditions of General License Issued Under § 72.210.” The term of a general 

license is tied to the term of the CoC being used. Within the general license term, each DSS

has its own storage term that begins when that DSS is placed into service at the ISFSI (see 

Appendix F for a discussion of storage terms). When a CoC is renewed, the associated users’ 

general licenses are also renewed. If the CoC holder chooses not to apply for the renewal of a 

particular CoC or is no longer in business, a licensee, licensee’s representative, or another 

certificate holder may apply for renewal of the CoC in place of the CoC holder.

Both the specific-license and the CoC renewal applications must contain requirements and 

operating conditions (fuel storage, surveillance and maintenance, and other requirements) for 

the ISFSI or DSS that address aging mechanisms and aging effects that could affect structures, 

systems, and components relied upon for the safe storage of spent fuel. Renewal applications 

must include (1) time-limited aging analyses, if applicable, that demonstrate that structures, 

systems, and components important to safety will continue to perform their intended function 

for the requested period of extended operation, and (2) aging management programs for 

management of issues associated with aging that could adversely affect structures, systems, 

and components important to safety. Licensees and applicants are encouraged to meet with 

the NRC staff at public pre-application meetings to discuss their proposed plans for the 

renewal application.

The technical review of the renewal application is primarily a materials engineering effort. The 

materials discipline should coordinate its review of the renewal application with other disciplines, 

such as the structural, radiation protection, thermal, criticality, and quality assurance disciplines, 

as appropriate, to help ensure that relevant aspects of the application and review have been 

addressed.

This SRP defines an acceptable method for the NRC staff to review and determine if the 

applicant demonstrates that the specific-licensed ISFSI or the certified DSS will continue to 

meet the applicable regulatory requirements of 10 CFR Part 72 during the period of extended 

operation. The reviewer should be aware that additional interim staff guidance may have been 

issued to clarify or address issues following the publication of this guidance. This SRP defines 
2

an acceptable method for satisfying the applicable regulatory requirements; it is not a regulatory 

requirement. An applicant may propose alternate means for satisfying the appropriate 

regulatory requirements. However, deviation from this guidance in whole or in part may result in 

an extended NRC staff review schedule.

The NRC expects to periodically revise and update this SRP to clarify the content, correct 

errors, and include new information, knowledge, and experience regarding aging management 

considerations. 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 Management, 

Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 

Washington, DC 20555-0001.

The guidance document is not intended to be used for the review of other 10 CFR Part 72 

renewal applications, such as those for wet storage facilities or monitored retrievable storage 

installations.

This guidance document is also not intended to be used as the sole guidance for the review of 

a specific license application for an “interim consolidated storage facility” (ICSF), where a DSS 

structure, system, and component (SSC) has been in storage at one location for some period of 

time and then is transported to a second location (ICSF) for subsequent storage. Guidance for 

review of an application for a 10 CFR Part 72 specific license is located in NUREG-1567, 

“Standard Review Plan for Spent Fuel Dry Storage Facilities.” However, as the DSS SSCs 

that will be in storage at a potential ICSF enter the period of extended operation, this guidance 

document is applicable to the aging management of such DSS SSCs.
3

Figure A is a flowchart of the specific-license and CoC renewal process.

Figure A. Specific-license and CoC renewal process

1. General Information Review

• General Information (1.4.1)

• Financial Information (1.4.2)

• Environmental Report (1.4.3)

• Application Content (1.4.4)

2. Scoping Evaluation

• Scoping Process (2.4.1)

• SSCs within the Scope of Renewal (2.4.2)

• SSCs not within the Scope of Renewal (2.4.3)

3. Aging Management Review

• Identification of Materials and Environments (3.4.1.1)

• Identification of Aging Mechanisms and Effects (3.4.1.2)

• Aging Management Activities (3.4.1.3)

• Aging Management Review for Fuel Assemblies (3.4.1.4)

3.5 Time-Limited Aging Analyses 3.6 Aging Management Programs
4

Revision 1

Based on lessons learned from reviews of specific-license and CoC renewal applications and 

input received from the public and industry, the NRC staff proposed changes to NUREG-1927, 

Revision 0, to add greater detail and clarity. The staff held public meetings, including a public 

meeting on July 14–15, 2014, to solicit stakeholder input on the staff’s considerations for 

revisions to the guidance. The staff subsequently took stakeholder input into consideration and 

developed the draft NUREG-1927, Revision 1, which was published for public comment on 

July 7, 2015 (80 FR 38780). The staff considered public comments received on the draft 

guidance in preparing the final NUREG-1927, Revision 1. The public comments are located in 

the Agencywide Documents Access and Management System (ADAMS) Package Accession 

No. ML15356A560. The staff also prepared responses to the public comments, at 

ADAMS Accession No. ML16125A534.

This revision of NUREG-1927 focuses on expanding guidance on application content, scoping 

evaluation, aging management review, time-limited aging analyses, and elements of an aging 

management program (AMP), including evaluation of AMPs and ensuring these programs 

respond to operating experience to remain adequate throughout the period of extended

operation (i.e., learning AMPs). This revision of NUREG-1927 also includes new guidance in 

the areas of: (1) timely renewal, (2) amendment applications submitted during renewal reviews 

and after the renewal is issued, (3) use of terms, conditions, and specifications for ensuring 

AMPs remain adequate during the period of extended operation, (4) commencement of AMPs 

for CoC renewals and implementation of AMPs, (5) example AMPs, (6) use of a demonstration 

program as a surveillance tool for high burnup fuel performance, and (7) storage terms (and 

calculation of length of time that a dry storage system can remain loaded).

This revision to NUREG-1927 was developed in parallel with an ongoing effort by the Nuclear 

Energy Institute (NEI) to develop guidance for the industry in the preparation of applications for 

renewal of specific licenses and CoCs. NEI 14-03, Revision 1, “Format, Content and 

Implementation Guidance for Dry Cask Storage Operations-Based Aging Management” 

(ADAMS Accession No. ML15272A329), includes guidance on the continued evaluation of 

operating experience (see Section 3.6.1.10 of NUREG-1927, Revision 1). One of the principles 

introduced in NEI 14-03 is the use of “tollgates” as a structured approach for assessing 

operating experience and data from applicable research and industry initiatives. In addition, 

NEI 14-03 describes an initiative to aggregate and disseminate aging-related operating 

experience, research results, monitoring feedback, and inspection data between licensees. 

The staff provided comments on NEI 14-03, Revision 0, “Guidance for Operations-Based Aging 

Management for Dry Cask Storage” (ADAMS Accession No. ML14266A225), to NEI on 

January 21, 2015 (ADAMS Accession No. ML15013A201). At the time of publication, the staff 

was continuing its review of NEI 14-03, Revision 1, for proposed NRC endorsement. However, 

until a time when NEI 14-03 may be endorsed by NRC, Section 3.6.1.10 of NUREG-1927, 

Revision 1, provides guidance to reviewers regarding information in NEI 14-03 that may be 

used or referenced by applicants for specific license or CoC renewals.
5

Standard Review Plan Structure

Each chapter of this SRP contains the following sections:

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

the major review objectives for the chapter. It also discusses the information needed, or 

coordination expected, from other NRC staff to complete the technical review.

Areas of Review: This section describes the structures, systems, and components; analyses,

data, or other information; and their sequence in the discussion of acceptance criteria.

Regulatory Requirements: This section summarizes the regulatory requirements in 

10 CFR Part 72 pertaining to the scoping process, aging management review, and aging 

management activities that include the time-limited aging analyses review. This list is not 

all-inclusive, and the reviewer should be aware that other parts of the regulations, such as 

10 CFR Part 20, “Standards for Protection against Radiation,” are assumed to apply to all 

licensees. 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.

Review Guidance: This section discusses the specific technical information that should be 

included in the application and reviewed for regulatory compliance. The review guidance can

be supplemented by interim staff guidance, NUREGs, etc.

Evaluation Findings: This section provides sample summary statements for evaluation findings 

to be incorporated into the safety evaluation report (SER) for each area of review. The reviewer 

prepares the evaluation findings based on the satisfaction of the regulatory requirements. The 

NRC publishes the findings in the SER.

7

1. GENERAL INFORMATION REVIEW

1.1 Review Objective

The purpose of the general information review is to ensure that the specific-license or certificate 

of compliance (CoC) renewal application meets the requirements listed in Section 1.3 below.

1.2 Areas of Review

Areas of review addressed in this chapter include the following:

• general information (specific license only)

• financial information (specific license only)

• environmental report (specific license only)

• application content

Areas specifically excluded from the renewal review include the following:

• structures, systems, and components (SSCs) associated with physical protection of the 

independent spent fuel storage installation (ISFSI) or dry storage system (DSS), under

Title 10 of the Code of Federal Regulations (10 CFR) Part 72, “Licensing Requirements 

for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste,

and Reactor-Related Greater Than Class C Waste,” Subpart H, “Physical Protection”

• SSCs associated with the ISFSI emergency plan, under 10 CFR 72.32, “Emergency 

Plan”

1.3 Regulatory Requirements

Table 1-1 presents a matrix that identifies the specific regulatory requirements pertaining to 

application content, general information about the specific licensee or CoC holder, financial 

information, and the environmental report. Additional regulatory requirements for the 

environmental report can be found in 10 CFR Part 51, “Environmental Protection Regulations 

for Domestic Licensing and Related Regulatory Functions.”

Table 1-1. Relationship of Regulations and General Information Review

Areas of Review

72.22 a

(a), (b), (c), (d)

72.22 a

(e) 72.30(c) a

Application Content ●

General Information ●

Financial Information ● ●

Environmental Report
8

Areas of Review

72.34 a 72.42 a 72.48(d) 72.240 b

(b), (c), (d)

Application Content ● ● ●

General Information

Financial Information

Environmental Report ●

a These regulations apply only to specific license renewals per 10 CFR 72.13, “Applicability.” b These regulations apply only to CoC renewals per 10 CFR 72.13.

1.4 Review Guidance

This section provides review guidance for general information (Section 1.4.1), financial 

information (Section 1.4.2), environmental report (Section 1.4.3), and application content 

(Section 1.4.4). This section also provides information on timely renewal (Section 1.4.5), 

amendment applications submitted during the renewal review or after the renewal is issued 

(Section 1.4.6) and license and CoC conditions (Section 1.4.7).

1.4.1 General Information

The U.S. Nuclear Regulatory Commission (NRC) project manager (PM) should ensure that the 

specific licensee has provided information under 10 CFR 72.22(a)–(d), including the specific

licensee’s full name, address, and description of the business or occupation. If the specific 

licensee is a partnership, the application should identify the name, citizenship, and address of 

each partner, and the principal location where the partnership does business. If the specific 

licensee is a corporation or an unincorporated association, the application should specify the 

State in which it is incorporated or organized and the principal location at which it does 

business, along with the names, addresses, and citizenships of its directors and principal 

officers. If the specific licensee is acting as an agent or representative of another person in 

filing the application, the application should provide the above information for the principal. If 

the specific licensee is the U.S. Department of Energy, then the application should specify the 

organization responsible for the construction and operation of the ISFSI and describe any 

delegations of authority and assignments of responsibilities.

1.4.2 Financial Information

In general, the PM should ensure that the renewal application for a specific license contains the 

necessary documentation regarding financial data, under 10 CFR 72.22(e), which shows that

the specific licensee can carry out the proposed activities for the requested duration. 

Information should state where the activity will be performed, the general plan for carrying out 

the activity, and the period of time for which the specific license is requested. The PM should 
9

ensure that the renewal application is based only on the approved design bases and does not 

include additional construction costs beyond the design bases. The application should identify 

other costs related to activities associated with managing aging mechanisms and effects, and it 

should identify ISFSI operating and decommissioning costs that have been revised from those 

specified in the original specific-license application for construction, operation, and 

decommissioning. In addition, the application should include a decommissioning funding plan 

that identifies any changes in decommissioning costs and the extent of contamination, pursuant 

to 10 CFR 72.30(c).

The scope of this standard review plan (SRP) does not include specific guidance for reviewing 

financial information. Financial reviews should be coordinated with financial reviewers in the 

Performance Assessment Branch of the Office of Nuclear Material Safety and Safeguards 

(NMSS) or the Office of Nuclear Reactor Regulation.

1.4.3 Environmental Report

The PM should ensure that the specific-license renewal application contains an environmental 

report or supplement, as required by 10 CFR 51.60, “Environmental Report—Materials 

Licenses” and 10 CFR 72.34, “Environmental Report.” The supplemental report may be limited 

to incorporating by reference, updating, or supplementing the information previously submitted 

to reflect any significant environmental changes, including those that may result from operating 

experience as related to environmental conditions, or a change in operations or proposed 

decommissioning activities.

The environmental report should also meet the general requirements of 10 CFR 51.45, 

“Environmental Report,” as applicable. As required by 10 CFR 51.45(c), the environmental 

report should contain sufficient data to aid the NRC in its development of an independent 

analysis.

The review of the environmental report should be coordinated with the Environmental Review 

Branch of NMSS.

1.4.4 Application Content

The reviewer should look for a map or guide to the renewal application to assist in its review, 

because the format may vary from that of a standard safety analysis report (SAR). The PM or 

reviewer should verify that the renewal application for both CoC and specific-license renewals 

contain all of the following sections:

• general Information (specific license only)

• scoping evaluation

• aging management review (AMR)

• time-limited aging analyses (TLAAs)

• aging management programs (AMPs)

• information pertaining to granted exemptions and their implication to aging management
10

• changes or additions to technical specifications or to the specific license or CoC

• supplement to the final safety analysis report (FSAR), including:

– scoping results

– table of AMR results

– summary of TLAAs and TLAAs’ conclusions

– summary of AMPs

• annotations to show 10 CFR 72.48 (“Changes, Tests, and Experiments”) changes since 

last biannual update as required by 10 CFR 72.48(d)(2) or any other changes from either 

the SAR pages included with the last approved amendment (or initial) application, or the 

FSAR, whichever is most recent

For CoC renewal applications that involve multiple amendments, the PM or reviewer should 

verify that the renewal application also includes:

• a description of the organization of the renewal application as it relates to the different 

amendments (i.e., for the CoC as a whole)

– This could be in the form of a guide to the renewal application to identify 

the sections or appendices applicable to each CoC amendment.

• a clear description of each amendment

– That is, what each amendment changed from (or added to) the initial certificate

(i.e., “amendment 0”), or what each amendment changed from (or added to) the 

previous amendments, should be described.

• a clear description of the scope and content of the renewal application as it applies to

each amendment

– If there are different SSCs, materials, contents specifications, or environments

described in the different CoC amendments, the application should specify any 

differences in the scoping evaluation, aging management review, TLAAs, and 

AMPs, for each individual amendment.

A CoC renewal includes the initial certificate (“amendment 0”) and all subsequent amendments. 

The subsequent amendments have the same termination date as the initial certificate. The CoC 

holder has the option to request that only certain (i.e., not all) amendments under a CoC be 

renewed. If amendments are not renewed, upon expiration, casks loaded under that 

amendment would need to be removed from service when they reach the end of their storage 

term (see Appendix F for calculation of storage terms). As a means to extend the storage term, 

a general licensee (cask user) also may have the option to apply changes authorized by another 

amendment to the CoC that has been renewed, to a cask loaded under an amendment that has 

not been renewed following the requirements in 10 CFR 72.212.

Drawings provided as part of the renewal application should be clear and legible. If information 

in drawings is unclear or illegible, the PM should ask the applicant for additional, larger or 

full-size drawings. The reviewer should ensure that dimensions, materials, and other details on 
11

the drawings are consistent with those described in both the text of the renewal application and 

the FSAR supplement.

The reviewer should verify that the applicant has updated the appropriate drawings to reflect 

any changes made to the design of the SSCs through the application of 10 CFR 72.48. 

Reviewers should be familiar with NUREG/CR-5502, “Engineering Drawings for 10 CFR Part 71 

Package Approvals,” issued May 1998. Although NUREG/CR-5502 was developed for 

transportation packages, the criteria for drawings are consistent for storage designs and 

therefore useful to the review process.

If the applicant provided drawings and descriptions as proprietary information in the application 

and requested them to be withheld from the public, the PM should review the request for 

withholding and ensure all the necessary information is available for the NRC to make a 

decision on the withholding request, in accordance with the requirements of 10 CFR 2.390, 

“Public Inspections, Exemptions, Requests for Withholding.” The applicant should also submit 

a nonproprietary version of the document to be made available to the public.

The reviewer should ensure the specific-license or CoC renewal application does not include 

any changes to the design bases. Changes to the design bases must be requested through a 

separate amendment process. However, the renewal application may include editorial changes 

or corrections that do not change the design bases.

1.4.5 Timely Renewal

To renew a specific license, an applicant must submit a renewal application at least 2 years 

before the expiration of the license in accordance with the requirements of 10 CFR 72.42(b). 

To renew a CoC, an applicant must submit a renewal application at least 30 days before the 

expiration of the CoC in accordance with the requirements of 10 CFR 72.240(b). When the 

applicant has submitted a timely application for renewal, the existing specific license or CoC will 

not expire until a final decision concerning the application for renewal has been made by the 

Commission. Therefore, any DSSs loaded during the initial license or CoC period may remain 

in service until the review of the renewal application is complete.

1.4.6 Amendment Applications Submitted during the Renewal Review or after 

the Renewal Is Issued

By regulation, applicants must demonstrate that SSCs important to safety will continue to 

perform their intended function(s) for the requested period of extended operation as a part of 

the renewal request. For concurrent amendment and renewal applications, the amendment 

application should include a scoping evaluation and an AMR for that amendment to document 

the evaluation of the amendment’s SSCs (and associated subcomponents) for extended 

operation, or the renewal application should be supplemented to address the proposed 

amendment to document the evaluation of the amendment’s SSCs (and associated 

subcomponents) for extended operation. Any amendment application submitted after the 

renewal has been issued (post-renewal amendment applications) should include a scoping 

evaluation and an AMR for that amendment. 

For post-renewal amendment applications or concurrent amendment applications that include 

a scoping evaluation and an AMR, the amendment application should either: (1) show that the

in-scope SSCs (and associated subcomponents) described in the amendment are already 

encompassed in the TLAAs or AMPs included in the specific-license or CoC renewal 
12

application, or (2) include revised or new TLAAs or AMPs to address aging effects of any new 

in-scope SSCs (and associated subcomponents) proposed in the amendment application. The 

PM and technical reviewers should verify that the following information is included in the 

amendment application (see also Section 1.4.4):

• a scoping evaluation that identifies any new SSCs (and associated subcomponents) 

included in the amendment request and discusses whether the SSCs (and associated 

subcomponents) are included or excluded from the scope of renewal, following the 

guidance in Chapter 2

• an aging management review that identifies any applicable aging mechanisms and 

effects for the new SSCs (and associated subcomponents) within the scope of renewal

• changes to the FSAR, which should include:

– scoping results and identification of any new in-scope SSCs

– revised table of AMR results

– identification of the approved TLAAs (or the TLAAs included in the renewal 

application, for concurrent amendments) that address the new in-scope SSCs,

or identification and a summary of any revised or new TLAAs and the TLAAs’ 

conclusions that support the amendment

– identification of the approved AMPs (or the AMPs included in the renewal 

application, for concurrent amendments) that encompass the new in-scope 

SSCs (and associated subcomponents), or a summary of proposed changes 

to approved AMPs (or the AMPs in the renewal application, for concurrent 

amendments) or new AMPs that will apply to the new in-scope SSCs (and 

associated subcomponents)

For concurrent amendment and renewal applications, if there are different PMs assigned to the 

renewal review and the amendment review, the PMs and technical reviewers should coordinate 

across the reviews to ensure that renewal aspects are covered for the amendment. Note that, 

before proceeding with the review of an amendment submitted during the renewal review, the 

PMs should consider how each review may affect the other, and decide, in conjunction with 

Branch and Division management, whether to proceed with both reviews, or to delay one review 

until the other is complete. For additional guidance, refer to Regulatory Issue Summary (RIS) 

2004-20, “Lessons Learned from Review of 10 CFR Parts 71 and 72 Applications, (NRC 2004).”

The NRC staff may include a condition in the renewed license or CoC noting all future 

amendments would need to address aging management.

1.4.7 Terms, Conditions, and Specifications for Specific Licenses and CoCs in the Period

of Extended Operation

In renewing a license or CoC, the staff should consider whether any terms, conditions, or 

specifications are needed to ensure the safe operation of the ISFSI or DSS during the period 

of extended operation, including but not limited to, terms, conditions, and specifications that will 

require implementation of any AMPs. Several conditions are likely to be included in the license 

or CoC.
13

Generally, NRC staff will renew the license or CoC with a condition requiring the specific

licensee or CoC holder, respectively, to incorporate a renewal supplement into the FSAR, 

as submitted in the renewal application and revised through the review process (see 

Section 1.4.4). The specific licensee will be required to continue to update the FSAR under the 

requirements in 10 CFR 72.70. The CoC holder will be required to continue to update the FSAR 

under the requirements in 10 CFR 72.248.

NRC staff may renew a license or CoC with a condition requiring licensees to implement 

the activities in the AMPs (e.g., update, revise, or create programs or procedures for AMP 

implementation) by a specific date (see Section 3.6.3). These programs and procedures will 

be subject to NRC inspection to ensure they are maintained, implemented, and periodically 

updated to respond to operating experience, while providing reasonable assurance that the 

pertinent SSCs will continue to perform their intended functions in the period of extended 

operation.

As the entirety of the AMP may not be included in the license, CoC, or technical specifications, 

site procedures for AMP implementation may later be changed without prior NRC review and 

approval. Therefore, the staff should consider whether additional conditions or specifications 

are needed in the specific license or CoC to ensure elements of the AMPs, which are the basis 

of the staff’s findings and the decision to issue the renewal, are effectively retained or 

implemented (e.g., timeframes for development of inspection or examination methods). These 

conditions should be specific to information in the AMP described in the renewal application 

which staff relied upon to make the requisite safety findings of reasonable assurance of 

adequate protection of public health and safety, and the environment.

The entire AMPs may be included as an appendix to the staff’s safety evaluation report (SER), 

for complete documentation of the staff’s safety evaluation, findings, and decision to issue the 

renewal. The staff should include a separate section in the SER that includes any new or 

modified terms, conditions, or specifications that were added to the license or CoC as a result

of the renewal review. This section should include the basis for each new or revised term, 

condition, or specification, or it should include a reference to the other sections of the SER 

that discuss the staff’s evaluation and findings that form the basis for the new or revised

term, condition, or specification.

For CoC renewals, additional terms, conditions, or specifications may be needed to ensure 

the safe operation of the cask during the renewal term, including but not limited to, terms, 

conditions, and specifications that will require the implementation of an AMP, in accordance with 

10 CFR 72.240(e). Such conditions may include requirements for a future or existing general 

licensee (cask user) to include in its 10 CFR 72.212(b)(5) evaluation (including the results of the 

review and determination per 10 CFR 72.212(b)(6) and 10 CFR 72.212(b)(8)) how it will meet 

the new CoC terms, conditions, or specifications for aging management (see Appendix E).

The NRC staff may also include a condition in the renewed license or CoC to ensure that all 

future amendments address the renewed design bases for the ISFSI or CoC, including any 

aging management considerations. See Section 1.4.6 for guidance on post-renewal 

amendments.

In addition, the applicant may also propose additional license or CoC conditions as part of its 

application, if these support the technical basis for the application and the proposed TLAAs or 

AMPs that ensure the safe operation of the ISFSI or DSS during the period of extended 

operation.
14

1.5 Evaluation Findings

The reviewer prepares a summary statement and evaluation findings based on compliance with

the regulatory requirements in Section 1.3. The summary statement and evaluation findings 

should be similar in wording to the following example (the finding number is for convenience in 

cross-referencing within the SRP and SER):

The NRC staff has reviewed the general information provided in the renewal application and 

supplemental documentation. The NRC staff performed its review following the guidance 

provided in NUREG-1927, Revision 1, “Standard Review Plan for Renewal of Specific Licenses 

and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel,” and relevant interim staff 

guidance. Based on its review, the NRC staff finds:

F1.1 The information presented in the renewal application satisfies the requirements of 

10 CFR 72.2, 72.22, 72.30, 72.34, 72.42, 72.48, and 72.240, as applicable.

F1.2 The applicant has provided a tabulation of all supporting information and docketed 

material incorporated by reference, in compliance with 10 CFR 72.42 or 72.240, 

as applicable.
15

2. SCOPING EVALUATION

2.1 Review Objective

The scoping evaluation should identify the structures, systems, and components (SSCs) of the 

independent spent fuel storage installation (ISFSI) or dry storage system (DSS) that should be 

reviewed for aging mechanisms and effects.

2.2 Areas of Review

The reviewer should ensure that the applicant has included information about the following 

areas of review:

• scoping process

• SSCs within the scope of specific-license or certificate of compliance (CoC) renewal

• SSCs not within the scope of specific-license or CoC renewal

2.3 Regulatory Requirements

The U.S. Nuclear Regulatory Commission (NRC) bases a specific-license or CoC renewal on 

the continuation of the approved design bases throughout the period of extended operation. 

The entire design bases of the specific license or CoC is considered to be renewed when NRC 

issues a renewed license or CoC. However, the NRC’s renewal review is focused on the 

maintenance of the intended functions of (a) SSCs important to safety and (b) SSCs failure of 

which may affect a safety function, as discussed in Section 2.4.2. This guidance document 

refers to such SSCs as those that are within the scope of renewal, and these are the SSCs that 

are further reviewed for aging mechanisms and effects, as discussed in Chapter 3, “Aging 

Management Review.”

If new safety-related deficiencies in the design bases are discovered, they must be addressed 

and rectified through the specific-license or CoC amendment process. The renewal process 

cannot be used to facilitate approval of design changes.

Table 2-1 presents a matrix of regulatory requirements for renewal related to the 

scoping review.
16

Table 2-1. Relationship of Regulations and Scoping Review

Areas of Review 10 CFR Part 72 Regulations

72.3 72.24 a

(b), (c), 

(d)

72.24 a

(g)

72.42 a

(b)

Subpart F: Applicable 

Sections

72.120, 72.122, 

72.124, 72.126, 72.128

72.236 b

Applicable

Sections

Scoping Process ● ● ●

SSCs within the 

Scope of 

Specific-License 

or CoC Renewal

● ● ● ● ●

SSCs not within the 

Scope of 

Specific-License 

or CoC Renewal

● ● ● ●

a These regulations apply only to specific-license renewals per Title 10 of the Code of Federal Regulations

(10 CFR) 72.13. b These regulations apply only to CoC renewals per 10 CFR 72.13.

2.4 Review Guidance

This section provides review guidance for the scoping evaluation. Section 2.4.1 explains the 

scoping process, Section 2.4.2 discusses SSCs within the scope of renewal, while Section 2.4.3 

provides guidance for SSCs not within the scope of renewal.

2.4.1 Scoping Process

Figure 2-1 provides a flowchart of the scoping evaluation process. The reviewer should ensure 

that the application provides documentation of the scoping process that includes the following:

• a description of the scoping process and method for the inclusion or exclusion of SSCs

(and associated subcomponents) from the renewal scope

• a list of the SSCs (and associated subcomponents) that are identified as within the 

scope of renewal, their intended function, and safety classification or basis for inclusion 

in the renewal scope

• a list of the SSCs (and associated subcomponents) that are identified as not within the 

scope of renewal and basis for exclusion

• a list of the sources of information used

• any discussion or drawings needed to clarify the process, SSC intended functions and 

safety classifications
17

Is the SSC 

important to 

safety?

Could the failure 

of the SSC not Important 

to Safety impact a 

safety function?

SSC is not

within the scope of

renewal

SSC is 

within the scope of

renewal 

(see Figure 3-1)

Yes

No

No

Yes

Scoping evaluation

process for

specific license or 

CoC renewal

Figure 2-1. Flowchart of scoping evaluation process

The application should include a list and description of reference sources used to support the 

scoping evaluation. Sources may include the following:

• safety analysis reports (SARs), including final SARs (FSARs), updated FSARs, and 

topical SARs

• license or CoC

• technical specifications

• approved exemptions

• operating procedures

• design-bases documents (e.g., calculations, specifications, design change documents)

• drawings
18

• quality assurance plan or program

• docketed correspondence

• operating experience reports (site-specific or industrywide, as applicable)

• Title 10 of the Code of Federal Regulations (10 CFR) 72.48 (“Changes, Tests, and 

Experiments”) evaluations and screenings

• vendor information

• applicable NRC guidance

The reviewer can refer to NUREG/CR-6407, “Classification of Transportation Packaging 

and Dry Spent Fuel Storage System Components According to Importance to Safety,” as a 

reference for classification of components as important to safety to determine the accuracy 

and completeness of the scoping evaluation. The reviewer should ensure that the scoping 

evaluation has evaluated all SSCs identified in the design-bases documents and properly 

differentiated them as either within or not within the scope of renewal. In addition, the 

identification of SSCs and SSC subcomponents in the scoping evaluation should be consistent 

throughout the application.

2.4.2 Structures, Systems, and Components within the Scope of Renewal

The reviewer should verify that the SSCs (and associated subcomponents) within the scope of 

renewal fall into the following scoping categories:

(1) They are classified as important to safety, as they are relied on to do one of the 

following functions:

i. maintain the conditions required by the regulations, specific license, or CoC to 

store spent fuel safely

ii. prevent damage to the spent fuel during handling and storage

iii. provide reasonable assurance that spent fuel can be received, handled, 

packaged, stored, and retrieved without undue risk to the health and safety of 

the public

These SSCs ensure that important safety functions are met for (1) confinement, 

(2) radiation shielding, (3) sub-criticality control, (4) heat-removal capability, (5) structural 

integrity, and (6) retrievability.

(2) They are classified as not important to safety but, according to the design bases, their 

failure could prevent fulfillment of a function that is important to safety.

The reviewer should verify that SSCs within the scope of renewal are screened to identify and 

describe the subcomponents with intended functions. The reviewer should recognize that SSC 

subcomponents may degrade by different modes, or have different criteria for evaluation from 

the overall component (i.e., different materials or environments).
19

The scoping evaluation should clearly (1) define the intended function of each SSC 

subcomponent and (2) differentiate SSC subcomponents per scoping criteria 1 and 2, as 

defined above. The reviewer should ensure that this information is tabulated or adequately 

described in the application. The reviewer should confirm that this information is 

comprehensive and accurate (i.e., SSC subcomponents are not missing from the scoping 

evaluation; SSC subcomponent naming is consistent with the design bases; intended functions 

are properly described) by comparing the results of the scoping evaluation to appropriate final 

safety analysis report (FSAR) drawings or tables.

2.4.2.1 Scoping of Fuel Assemblies

Traditional light water reactor spent nuclear fuel consists of fuel rods and assembly hardware. 

In turn, the fuel rods consist of uranium oxide pellets inside a cladding tube. The spent fuel 

cladding and assembly hardware provide structural support to ensure that the spent fuel is 

maintained in a known geometric configuration. The safety analyses for an ISFSI or DSS 

(e.g., criticality and shielding analyses) may rely on the fuel assembly having a specific 

configuration (e.g., geometric form, a certain number of fuel rods or solid replacement filler rods 

in the assembly lattice). As the renewal of a specific license or CoC is based on continuation of 

the approved design bases throughout the period of extended operation (as discussed in 

Section 2.3), for these ISFSIs or DSSs, the renewal application should demonstrate that the 

analyzed fuel configuration is maintained during the period of extended operation. Therefore, 

the condition of the fuel assembly and cladding are within the scope of renewal and should be 

reviewed for any aging mechanisms and effects that may lead to a change in the analyzed fuel 

configuration. If a licensee or CoC holder wishes to revise the safety analyses for the approved 

design-bases fuel configuration (i.e., analyzed fuel configuration), it should pursue such a 

change through an amendment or revision request, and not as part of the renewal application.

2.4.2.2 Scoping of Structures, Systems, and Components, Depending on Individual Design 

Bases

In some cases, transfer casks, transporter devices, reinforced concrete pads, and other 

engineered features (e.g., earthen berms, shield walls, or engineered fill within an underground 

ISFSI) may be classified as important to safety or safety-related (under 10 CFR Part 50) in the 

design bases of various ISFSIs or DSSs. The reviewer should review the FSAR to determine 

how these SSCs are used in the FSAR evaluations and described in the license or CoC, to 

understand whether these SSCs are considered part of the design bases, and thus whether 

they are considered to be within the scope of renewal.

2.4.3 Structures, Systems, and Components Not within the Scope of Renewal

For those SSCs (and associated subcomponents) excluded from the scope of renewal, the 

reviewer should verify that they do not meet either of the criteria described in Section 2.4.2. 

The reviewer should ensure that the applicant has properly justified any exclusions 

by referencing the design bases (i.e., FSAR description, drawings, or tables).

The following SSCs may be excluded from the scope of renewal, provided that they do not meet 

either of the criteria in Section 2.4.2 above:

• equipment associated with cask loading and unloading, such as (1) welding and sealing 

equipment, (2) lifting rigs and slings, (3) vacuum-drying equipment, (4) portable radiation 
20

survey equipment, and (5) other tools, fittings, hoses, and gauges associated with cask 

loading and unloading

• Instrumentation and other active components/systems (i.e., not passive or long-lived, but 

subject to a change in configuration or replacement based on a qualified life or service 

time period)

• miscellaneous hardware that does not support or perform any function that is important

to safety

2.5 Evaluation Findings

The reviewer prepares the summary statement and evaluation findings based on compliance 

with the regulatory requirements described in Section 2.3. The summary statement and 

evaluation findings should be similar in wording to the following example (the finding number 

is for convenience in cross-referencing within the Standard Review Plan (SRP) and safety 

evaluation report (SER)):

The NRC staff reviewed the scoping evaluation provided in the renewal application and 

supplemental documentation. The NRC staff performed its review following the guidance 

provided in NUREG-1927, Revision 1, “Standard Review Plan for Specific Licenses and 

Certificates of Compliance for Dry Storage of Spent Nuclear Fuel,” and relevant ISGs. The 

NRC staff used the information provided in NUREG/CR-6407 (“Classification of Transportation 

Packaging and Dry Spent Fuel Storage System Components According to Importance to 

Safety”) in its review as a reference for classification of components as important to safety to 

determine the accuracy and completeness of the scoping evaluation. Based on its review, the 

NRC staff finds:

F2.1 The applicant has identified all SSCs important to safety and SSCs failure of which could 

prevent an SSC from fulfilling its safety function, per the requirements of 10 CFR 72.3, 

10 CFR 72.24, 10 CFR 72.42, 10 CFR 72.120, 10 CFR 72.122, 10 CFR 72.124, 

10 CFR 72.126, 10 CFR 72.128 and 10 CFR 72.236, as applicable.

F2.2 The justification for any SSC determined not to be within the scope of the renewal is 

adequate and acceptable.
21

3. AGING MANAGEMENT REVIEW

3.1 Review Objective

The purpose of the aging management review (AMR) is to assess the proposed aging 

management activities (AMAs) for structures, systems, and components (SSCs) determined to 

be within the scope of renewal. The AMR addresses aging mechanisms and effects 1 that could 

adversely affect the ability of the SSCs (and associated subcomponents) from performing their

intended functions during the period of extended operation. The reviewer should verify that the 

renewal application includes specific information that clearly describes the AMR performed on 

SSCs within the scope of renewal.

3.2 Areas of Review

The reviewer should ensure that the AMR in the renewal application provides the following 

content with adequate technical bases:

• identification of materials and environments for those SSCs and 

associated subcomponents determined to be within the scope of renewal

• identification of aging mechanisms and effects requiring management

• identification of time-limited aging analyses (TLAAs), if applicable, and

aging management programs (AMPs) for managing the effects of aging

Figure 3-1 contains a flowchart for the AMR process. The final safety analysis report (FSAR)

and supporting documents related to the design are the primary documents that describe the 

safety classification, intended function, materials, and service environments for SSCs of 

independent spent fuel storage installations (ISFSIs), dry storage systems (DSSs), or both, 

identified to be within the scope of renewal (see Section 2.4.1).

The reviewer should consult applicable consensus codes and standards that provide additional 

guidance on the applicability of aging mechanisms and effects. The use of ambiguous 

terminology from any standard (e.g., “change in material properties”) should be properly defined 

or referenced in the application. Refer to Appendix A for assessing non-quantifiable terms.

 1 To effectively manage an aging effect, it is necessary to determine the aging mechanisms that are 

potentially at work for a given material and environment application. Therefore, the aging management 

review process identifies both the aging effects and the associated aging mechanisms that cause them.
22

SSC or SSC subcomponent

within the scope 

of renewal (2.4.2)

Identification of 

materials and 

service environments

(3.4.1.1)

Determination of the 

AMA required to manage

the effects of aging 

(3.4.1.3)

Cont.

on next

page

Is the SSC or 

SSC subcomponent

subject to an 

aging mechanism 

or effect

(3.4.1.2)?

Document in 

FSAR supplement

or in the application

No

Yes

Figure 3-1. Flowchart of AMR process
23

Is the TLAA 

reconfirmed for the 

renewal period?

TLAA confirmed and

documented in FSAR 

supplement.

No further action 

necessary

Yes

Yes

No Can TLAA 

adequately predict 

degradation associated 

with identified 

aging effect?

No

Action Required:

Modification of an AMP 

in Appendix B or

introduction of a

new AMP to be 

reviewed by the NRC

and documented in

FSAR supplement

TLAA AMP

Cont.

from

previous

page

Figure 3-1. Flowchart of AMR process (continued)

3.3 Regulatory Requirements

Title 10 of the Code of Federal Regulations (10 CFR) 72.42 and 10 CFR 72.240 provide the 

overarching requirements for aging management activities for renewal of specific licenses and 

certificates of compliance (CoCs). Table 3-1 presents a matrix of regulatory requirements that 

must continue to be met to ensure that the intended functions of SSCs of ISFSIs and DSSs are 

maintained during the period of extended operation. Other parts of 10 CFR Part 72 may also 

apply.
24

Table 3-1. Relationship of Regulations and Aging Management Reviews

Areas of Review 10 CFR Part 72 Regulations

72.24a

(d)

72.82 a

(d)

72.104 a

(a)

72.106 a

(b)

72.120 a

(a),(d)

Aging Effects ● ● ● ●

Aging Management, Maintenance, 

or Surveillance Programs ●

TLAAs ● ● ● ●

Areas of Review 10 CFR Part 72 Regulations

72.122 a

(a),(b),(c),(h)(1),

(h)(5),(l)

72.122 a

(f),(h)(4),(i)

72.124 72.126a 72.128 a

(a)

Aging Effects ● ● ●

Aging Management, Maintenance, 

or Surveillance Programs ● ● ●

TLAAs ● ● ● ● ●

Areas of Review 10 CFR Part 72 Regulations

72.158 72.162 72.164 72.168

(a)

72.170

Aging Effects ● ● ●

Aging Management, Maintenance, 

or Surveillance Programs ● ● ● ● ●

TLAAs ●
25

Table 3-1. Relationship of Regulations and Aging Management Reviews (continued)

Areas of Review 10 CFR Part 72 Regulations

72.172 72.236 b

Applicable 

Sections

72.240(d) b

Aging Effects ● ●

Aging Management, Maintenance, 

or Surveillance Programs ● ● ●

TLAAs

a These regulations apply only to specific-license renewals per 10 CFR 72.13. b These regulations apply only to CoC renewals per 10 CFR 72.13.

3.4 Materials, Service Environments, Aging Mechanisms and Effects, and Aging 

Management Activities

3.4.1 Review Guidance

This section provides review guidance for the aging management review. Section 3.4.1.1 

describes the identification of materials and their environments. Section 3.4.1.2 describes the 

identification of aging mechanisms and effects. Section 3.4.1.3 describes the review of aging 

management activities. Section 3.4.1.4 describes the aging management review for spent 

fuel assemblies.

3.4.1.1 Identification of Materials and Environments

The AMR process includes the identification of the materials of construction and the service 

environments for each SSC (or SSC subcomponent) within the scope of renewal. The 

identification of SSCs and SSC subcomponents in the AMR process should be consistent 

with the identification in the scoping evaluation. The reviewer should ensure that the renewal 

application has provided environmental data (and referenced the source of the data) such that 

the range of operating and service conditions of the SSCs can be determined. The pertinent 

environmental data is that which has a direct bearing on aging and the proposed aging 

management approach, and may include:

• temperature

• wind

• relative humidity

• relevant atmospheric pollutants and deposits

• exposure to precipitation

• marine fog, salt, or water exposure

• radiation field (gamma and neutron)

• the service environment (e.g., embedded, sheltered, or outdoor)

• gas compositions (e.g., external: air; internal: inert gas such as helium)
26

The reviewer should verify that the applicant considered the specific environment and material 

combinations in the DSS design for the evaluation of observed and potential aging mechanisms 

and effects that may lead to a loss of intended function. Particular attention should be paid to 

dissimilar metal combinations that may result in galvanic effects such as hydrogen uptake or 

galvanic corrosion. For sheltered environments that are in contact with the site atmosphere, 

the potential deposition and accumulation of atmospheric deposits should be evaluated. For 

example, chloride salt deposits that can promote accelerated corrosion of SSCs may be 

relevant in locations near salt water, adjacent to cooling towers, or in close proximity to roads 

that are treated with deicing salts.

3.4.1.2 Identification of Aging Mechanisms and Effects

The reviewer should ensure that the applicant has provided an analysis and documentation that 

identify aging mechanisms and effects pertinent to the SSCs determined to be within the scope 

of renewal. The AMR should include aging mechanisms and effects that could reasonably be 

expected to occur, as well as those that have actually occurred, based on industry and 

site-specific operating experience and component testing. The reviewer should ensure that the 

CoC renewal evaluates all potential environments where the DSS may be used and provides 

applicable operating experience to justify the AMR conclusions.

The reviewer should review the applicant’s synopses of information used to identify applicable 

aging mechanisms and effects. Identification of applicable aging mechanisms and effects may 

be through review of:

• site maintenance, repair, and modification records

• corrective action reports, including root cause evaluations

• pre-application inspection results (see below)

• maintenance and inspection records from ISFSI sites with similar SSC materials and 

operating environments

• industry records

• applicable operating experience outside the nuclear industry

• applicable consensus codes and standards

• U.S. Nuclear Regulatory Commission (NRC) reports

• other applicable guidance for determining if an aging mechanism or effect should be

managed for the period of extended operation

Examples of potential aging mechanisms and effects that may be identified by reviewing the 

sources of information cited above include: (1) cracking or loss of strength as a result of cement 

aggregate reactions in the concrete, (2) cracking or loss of material as a result of freeze-thaw 

degradation of the concrete (requires the presence of moisture combined with temperatures 

below freezing), (3) reinforcement corrosion and concrete cracking as a result of chloride 

ingress, (4) accelerated corrosion of steel structures and components and stress corrosion 

cracking of austenitic stainless steels as a result of atmospheric deposition of chloride salts. 
27

The reviewer should ensure that, if the applicant relies on operating experience, it is specifically 

applicable to the SSC subcomponent/material/environment, and is not just a compendium of 

references on similar topics.

The applicant is not required to take further action if an SSC is determined to be within the 

scope of renewal but is found to have no potential aging effects for the period of extended 

operation. The reviewer should verify that the applicant’s exclusion of an aging mechanism or 

effect is consistent with maintenance records, operating experience, and information obtained 

during pre-application inspection(s). The reviewer should also ensure that the FSAR 

supplement or other application materials document the applicant’s determination of SSCs 

requiring no further review.

Pre-Application Inspections

Because inspections of DSSs are not typically conducted during the initial storage period, or 

are narrow in scope, the reviewer may have limited information regarding applicable aging 

mechanisms and effects for the specific design, environment, and operational parameters. 

Although there is no specific regulatory requirement for a pre-application inspection, it is one 

means by which an applicant can demonstrate that an aging effect does or does not require 

management. It can provide valuable operating experience and can be used to verify the 

condition of SSCs and SSC subcomponents is as-expected. The pre-application inspection 

is performed before submittal of the specific-license or CoC renewal application, and the 

inspection results become part of the technical bases for renewal.

Pre-application inspections should bound the site conditions, system designs, material 

combinations, and operating parameters that may contribute to the potential aging mechanisms 

and effects for SSCs and SSC subcomponents within the scope of renewal. For example, if 

chloride-induced stress corrosion cracking (CISCC) of a stainless steel canister is a potential 

aging mechanism, then the pre-application inspection should be conducted on the canister that 

has the greatest susceptibility to CISCC. The determination of susceptibility may involve the 

initial heat load of the canister, expected temperature variations on the canister surface with 

priority given to the coolest locations and welds, and the canister location if it is determined that 

some DSSs at a site may be located closest to or oriented toward a source of atmospheric 

chlorides. To address such variables, particularly for CoC renewal applications, pre-application 

inspections may involve SSCs in multiple DSSs at an ISFSI, and DSSs at multiple ISFSIs, as 

applicable.

Pre-application inspections may not include transfer casks or other similar SSCs that are 

leased or otherwise not actually on site. The latter SSCs are generally subject to maintenance 

requirements before use. Records from these maintenance activities may be included in the 

application in support of their respective aging management programs. Applicants are 

encouraged to discuss their considerations for selecting the system(s) to inspect with NRC 

staff in pre-application meetings before submitting the renewal application.

The reviewer should ensure that the scope, methods, and acceptance criteria for pre-application 

inspections align with the guidance for the AMP elements described in Section 3.6.1 of this 

document. The reviewer may use technical information in the example aging management 

programs in Appendix B as a reference when reviewing the adequacy of pre-application 

inspections. The reviewer should ensure that the application provides a description of any 

initiated corrective actions (including results from actions to verify extent of condition) due to

conditions identified in the pre-application inspection(s).
28

The reviewer may accept the use of surrogate inspections (inspections conducted at other sites 

as a substitute for inspections conducted at the site(s) within the subject license or CoC) for 

identifying the relevant aging mechanisms and effects in the renewal application, but only when 

the technical basis is supported by substantial operating experience. Differences in materials, 

fabrication practices, design modifications, and environmental conditions at various sites could 

make comparisons between different ISFSI sites invalid.

The CoC holder is responsible for providing the technical basis for the proposed approach 

to aging management at multiple sites (where the CoC is and can be used). Therefore, 

pre-application inspections are likely only practical at a subset of sites that use the CoC. In this 

case, the reviewer should ensure that the chosen subset of sites is bounding with respect to the 

susceptibility of the various potential aging effects. Although CoC holders do not have the 

authority to conduct pre-application inspections at general licensee ISFSI sites, CoC holders 

could work within their user groups to identify bounding systems for the pre-application 

inspections. If pre-application inspections are not performed for each of the sites with the 

subject license or CoC, the reviewer should ensure that aging management programs include 

baseline inspections at each site upon entering the period of extended operation.

During baseline inspections, or the first inspections conducted per the AMPs, each licensee 

should assess the condition of SSCs: (1) to confirm the results of pre-application inspections 

that were conducted at other sites are bounding, or (2) to verify the adequacy of the AMPs and 

the conclusions of the TLAAs, when pre-application inspections were not performed. Also, the 

reviewer should consider whether uncertainties in SSC degradation due to the lack of a 

pre-application inspection warrant conditions in the license or CoC to require baseline

inspections immediately upon entering the period of extended operation (see Section 1.4.7).

3.4.1.3 Aging Management Activities

The reviewer should ensure that the applicant has identified those aging mechanisms and 

effects requiring either an AMP or TLAA. Figure 3-1 illustrates the process for handling those 

SSCs that are determined to be within the scope of renewal and subject to a potential aging 

effect. The AMR defines two methods for addressing potential aging mechanisms and effects: 

TLAA (Section 3.5) and AMP (Section 3.6).

The NRC may condition the approval of a renewal on the requirements of a given AMP being 

met during the period of extended operation (see Section 1.4.7). The CoC user (general 

licensee) would ordinarily carry out the activities described in this AMP. Under

10 CFR 72.212(b)(11) and 10 CFR 72.240(e), the NRC may add the appropriate condition(s) or 

technical specification(s) to the renewed CoC for general licensee implementation of the AMP. 

Specific licenses may also be similarly conditioned (see Section 1.4.7).

3.4.1.4 Aging Management Review for Fuel Assemblies

Because the DSS interior and cladding cannot be reasonably inspected, the reviewer should 

rely on lessons learned from NUREG/CR-6745, “Dry Cask Storage Characterization Project—

Phase 1; CASTOR V/21 Cask Opening and Examination,” (Bare et al., 2001), and 

NUREG/CR-6831, “Examination of Spent PWR Fuel Rods after 15 Years in Dry Storage,” 

(Einziger et al., 2003). This research demonstrated that low burnup fuel cladding and other 

cask internals had no deleterious effects after 15 years of storage. This research confirmed 

the basis for the guidance on creep deformation in Interim Staff Guidance (ISG) 11, “Cladding 

Considerations for the Transportation and Storage of Spent Fuel,” Revision 3, (NRC 2003). The 
29

NRC staff has indicated in ISG 11, Revision 3 that the spent fuel configuration is expected to be 

maintained as analyzed in the safety analyses for the ISFSI or DSS, provided certain 

acceptance criteria (regarding maximum fuel clad temperature and thermal cycling) are 

met, and the fuel is stored in a dry inert atmosphere. These research results suggest that 

degradation of low burnup fuel cladding and assembly hardware should not occur during the 

first renewal period, provided that the cask or canister internal environment is maintained.

The reviewer should assess whether the applicant has considered the most recent revision of 

ISG-11 and research results in this area, especially with respect to high burnup (HBU) fuel. 

Research into fuel performance in storage is an ongoing effort.

2

 The reviewer should ensure 

that the applicant has monitored new research developments to ensure it has identified any 

new potential aging mechanisms and effects and provided new supporting data demonstrating 

HBU fuel performance during the period of extended operation. Although NRC has confidence 

based on short-term testing (i.e., laboratory scale testing up to a few months) that there is no 

degradation of HBU fuel that would result in spent fuel being in an unanalyzed configuration in 

the period of extended operation, there is no operational data to demonstrate this as was done 

in the aforementioned demonstration on low burnup fuel.

Guidance for one acceptable approach to demonstrate the HBU fuel performance during the 

period of extended operation is provided in the “Example of a High Burnup Fuel Monitoring and 

Assessment Program” in Appendix B. This is a licensee program that monitors and assesses 

data and other information regarding HBU fuel performance, to confirm that the analyzed HBU 

fuel configuration is maintained during the period of extended operation. Guidance for 

determining if a surrogate demonstration program can provide the data to support a licensee’s 

HBU Fuel Monitoring and Assessment Program is given in Appendix D. Alternative 

approaches to that presented in Appendix B may be used, as appropriately justified by 

an applicant.

3.4.2 Evaluation Findings

The reviewer prepares the summary statement and evaluation findings based on compliance 

with the regulatory requirements in Section 3.3. The summary statement and evaluation 

findings should be similar in wording to the following example (the finding number is for 

convenience in cross-referencing within the Standard Review Plan (SRP) and safety evaluation 

report (SER)):

The NRC staff reviewed the aging management review provided in the renewal application 

and supplemental documentation. The NRC staff performed its review following the guidance 

provided in NUREG-1927, Revision 1, “Standard Review Plan for Renewal of Specific Licenses 

and Certificates of Compliance of Dry Storage of Spent Nuclear Fuel,” and relevant ISGs. 

Based on its review, the NRC staff finds:

F3.1 The applicant’s AMR process to be comprehensive in identifying the materials of 

construction and associated operating environmental conditions for those SSCs within the 

scope of renewal and has provided a summary of the information in the renewal application 

and FSAR supplement.

 2 For example, research programs at Argonne National Laboratory (for NRC and the U.S. Department of 

Energy (DOE)), and other programs performed for the Nuclear Regulation Authority (formerly Japan Nuclear 

Energy Safety Organization) have studied hydride reorientation effects; Oak Ridge National Laboratory has 

also studied bending responses of the fuel.
30

F3.2 The applicant’s AMR process to be comprehensive in identifying all pertinent aging 

mechanisms and effects applicable to the SSCs within the scope of renewal and has provided 

a summary of the information in the renewal application and FSAR supplement.

3.5 Time-Limited Aging Analyses

TLAAs are calculations or analyses used to demonstrate that in-scope SSCs will maintain their 

intended function throughout an explicitly stated period of extended operation (e.g., 40 years). 

These calculations or analyses may be used to assess fatigue life (number of cycles to 

predicted failure), or time-limited life (operating timeframe until expected loss of intended 

function). TLAAs should account for environmental effects. Under 10 CFR 72.3, “Definitions,” 

TLAAs are those calculations and analyses meeting all six of the following criteria:

(1) Involve SSCs important to safety within the scope of the specific-license renewal, as 

delineated in Subpart F of 10 CFR Part 72, or within the scope of the spent fuel storage 

CoC renewal, as delineated in Subpart L of 10 CFR Part 72, respectively.

(2) Consider the effects of aging.

(3) Involve time-limited assumptions defined by the current operating term.

(4) Were determined to be relevant by the specific licensee or certificate holder in making a 

safety determination.

(5) Involve conclusions or provide the basis of conclusions related to the capability of SSCs 

to perform their intended safety functions.

(6) Are contained or incorporated by reference in the design bases.

Under 10 CFR 72.42(a)(1) or 72.240(c)(2), the reviewer should ensure that the application 

includes a list of TLAAs. The NRC staff uses the FSAR and other documents where the design 

bases are detailed to perform the review and confirm that the applicant did not omit any TLAAs 

submitted as part of the approved design bases. The number and type of TLAAs vary 

depending on the design bases of the ISFSI or DSS. The reviewer should ensure that all 

six criteria set forth in 10 CFR 72.3 (and repeated in this section) are satisfied to conclude that a 

calculation or analysis is a TLAA.

The following examples illustrate analyses that are not TLAAs and need not be addressed 

under 10 CFR 72.42(a)(1) or 72.240(c)(2):

• Analyses with time-limited assumptions defined short of the initial license or CoC term, 

for example, an analysis for a component based on a service life that would not reach 

the end of the initial license or CoC term.

• Analyses not contained or incorporated by reference in the design bases. Although not 

TLAAs by definition, the reviewer should note that these analyses may be included in the 

renewal application to justify the proposed attributes of an AMP (e.g., inspection 

frequency, sample size) for a particular SSC within the scope of renewal. These 

analyses may further be used in the AMR to justify the exclusion of an aging 

mechanism/effect or SSC subcomponent from the scope of an AMP (i.e., not requiring 
31

any aging management activities), if the analysis is approved by the staff and included 

or summarized in the application.

3.5.1 Review Guidance

The reviewer should ensure that the applicant has appropriately identified TLAAs by applying 

the six criteria described below for SSCs within the scope of renewal:

(1) Involve SSCs important to safety within the scope of the specific-license or CoC 

renewal. Chapter 2 of this SRP provides the reviewer guidance on the scoping 

methodology.

(2) Consider the effects of aging. The effects of aging include but are not limited to loss 

of material, change in dimension, change in material properties, loss of strength, 

settlement, and cracking. The reviewer should ensure that any calculations or analyses 

relying on environmental susceptibility criteria are adequately supported by a valid 

technical basis, such as NRC endorsed criteria or operating experience. An AMP might 

be more applicable in these cases.

(3) Involve time-limited assumptions defined by the current operating term. The defined 

operating term should be explicit in the analysis. Simply asserting that the SSC is 

designed for a DSS or ISFSI service life is not sufficient. Calculations, analyses, or 

testing that explicitly include a time limit should support the assertions.

(4) Were determined to be relevant by the licensee or certificate holder in making a safety 

determination. Relevancy is a determination that the applicant makes based on a review 

of the information available. A calculation or analysis is relevant if it can be shown to 

have a direct bearing on the action taken as a result of the analysis performed. 

Analyses are also relevant if they provide the basis for a safety determination, and, 

in the absence of analyses, the applicant might have reached a different conclusion.

(5) Involve conclusions or provide the basis of conclusions related to the capability of SSCs 

to perform their intended safety functions. The TLAA should provide conclusions or a 

basis for conclusions regarding the capability of the SSC to perform its intended function 

through the end of the period of extended operation. If the TLAA does not provide a 

conclusion supporting the capability of the SSC to perform its intended function through 

the end of the period of extended operation, then the TLAA is not confirmed, and the 

applicant should propose an AMP to address/manage the aging mechanism or effect 

on the SSC. Analyses that do not affect the intended functions of SSCs are not TLAAs.

(6) Are contained or incorporated by reference in the design bases. TLAAs should already 

be contained or incorporated by reference in the design-bases documents. Such 

documentation includes the (1) FSAR, (2) technical specifications, (3) correspondence 

to and from the NRC, (4) quality assurance plan, and (5) topical reports included as 

references in the FSAR. The reviewer should ensure that the applicant has provided 

any references cited in design-bases documents that may be needed to clarify the 

assumptions, methods, or values used in TLAAs.

The reviewer should ensure that the applicant has appropriately dispositioned an identified 

TLAA by one of the following methods:
32

• Demonstrate the existing analysis remains valid for the period of extended operation, 

has already considered the requested period of extended operation, and concludes that 

the SSC will continue to perform its intended function through the end of the requested 

period of extended operation.

• Revise or update the existing analysis to demonstrate that it has been projected to the

end of the requested period of extended operation and concludes that the SSC will 

continue to perform its intended function through the end of the requested period of 

extended operation.

• Manage the effects of aging on the SSC for the requested period of extended operation

through an AMP.

3.5.2 Evaluation Findings

The reviewer prepares the summary statement and evaluation finding based on compliance 

with the regulatory requirements in Section 3.3. The summary statement and evaluation finding 

should be similar in wording to the following example (the finding number is for convenience in 

cross-referencing within the SRP and SER):

The NRC staff reviewed the TLAAs provided in the renewal application and supplemental 

documentation. The NRC staff performed its review following the guidance provided in 

NUREG-1927, Revision 1, “Standard Review Plan for Renewal of Specific Licenses and 

Certificates of Compliance for Dry Storage of Spent Nuclear Fuel,” and relevant ISGs. The 

NRC staff verified that the TLAA assumptions, calculations, and analyses are adequate and 

bound the environment, and aging mechanism or aging effect for the pertinent SSCs. Based 

on its review, the NRC staff finds:

F3.3 The applicant identified all aging mechanisms and effects pertinent to SSCs within the 

scope of renewal that involve TLAAs. The methods and values of the input parameters for the 

applicant’s TLAAs are adequate. Therefore, the applicant’s TLAAs provide reasonable 

assurance that the SSCs will maintain their intended function(s) for the period of extended 

operation, require no further aging management activities, and meet the requirements in 

10 CFR 72.42(a)(1) or 10 CFR 72.240(c)(2), as applicable.

3.6 Aging Management Programs

AMPs monitor and control the degradation of SSCs within the scope of renewal so that aging 

effects will not result in a loss of intended functions during the period of extended operation. An 

AMP includes all activities that are credited for managing aging mechanisms or effects for 

specific SSCs, including activities conducted during the initial storage period. An effective AMP 

prevents, mitigates, or detects the aging effects and provides for the prediction of the extent of 

the effects of aging and timely implementation of corrective actions before there is a loss of 

intended function.

AMPs should be informed, and enhanced when necessary, based on the ongoing review of 

both site-specific and industrywide operating experience, including relevant international and 

non-nuclear operating experience. Operating experience provides direct confirmation of the 

effectiveness of an AMP and critical feedback for the need for improvement. As new knowledge 

and data become available from new analyses, experiments, and operating experience, 
33

licensees and CoC holders should revise existing AMPs (or pertinent procedures for AMP 

implementation) to address program improvements or aging issues.

3.6.1 Review Guidance

An AMP should contain the following 10 elements:

(1) scope of program

(2) preventive actions

(3) parameters monitored or inspected

(4) detection of aging effects

(5) monitoring and trending

(6) acceptance criteria

(7) corrective actions

(8) confirmation process

(9) administrative controls

(10) operating experience

Review of the AMPs should include an assessment of the 10 program elements to verify their 

technical adequacy. In general, the reviewer should examine the details of these 10 elements 

for managing the aging mechanisms and effects identified by the aging management review 

(AMR) process. The reviewer should recognize that an applicant may develop AMPs following 

a different format or style. For such reviews, the NRC staff should ensure that sufficient detail 

(i.e., supporting technical bases) is provided in the alternate format in comparison with the 

10 AMP elements of this guidance.

The reviewer should determine if the proposed AMP is adequate for managing the aging

mechanisms and effects of the SSCs identified by the AMR. The following sections provide 

specific guidance for the review of each element of an AMP.

3.6.1.1 Scope of Program

The scope of the program should list the specific SSCs and subcomponents covered by the 

AMP and the intended functions to be maintained. In addition, the element should state the 

specific materials, environments, and aging mechanisms and effects to be managed. The 

reviewer should verify that the scope defined for the AMP is clear and specific.

3.6.1.2 Preventive Actions

Preventive actions are used to prevent aging or mitigate the rates of aging for SSCs through

the activities in the AMP. The reviewer should verify that these activities, if applicable, are 

described. For example, an applicant may cite a ground dewatering system to ensure control of 

long-term settlement of structures or the continuance of inspections to ensure that air inlet/outlet 

vents are not blocked. Some condition or performance monitoring programs do not rely on 

preventive actions and thus this information need not be provided.

The reviewer should ensure that any proposed preventive action will not result in an unintended 

consequence to the ability of an SSC to fulfill its intended function(s). For example, if the 

applicant has proposed to change a coating system to prevent loss of material due to corrosion, 

the reviewer should ensure that the applicant has verified coating compatibility to confirm that 

the proposed action will not compromise the intended function(s) of the SSC.
34

3.6.1.3 Parameters Monitored or Inspected

This program element should identify the specific parameters that will be monitored or inspected 

and describe how those parameters will be capable of identifying degradation or potential 

degradation before a loss of intended function. The use of the parameters should be 

demonstrated to be capable of:

• monitoring the effectiveness of activities that prevent or mitigate aging 

(e.g., environmental controls)

• monitoring the performance of SSCs as an indirect indicator of degradation 

(e.g., radiation rate monitoring at the external surface of a cask)

• detecting, through direct inspection, the presence and severity of conditions or 

discontinuities that may have an effect on the function of SSCs (e.g., nondestructive 

examination of a component surface)

The reviewer should ensure that this program element provides a clear link between the aging 

effects identified in the scope of the program and the parameters monitored or inspected.

3.6.1.4 Detection of Aging Effects

Detection of aging effects should occur before there is a loss of intended function for any 

SSC identified within the scope of the program. This element should include inspection and 

monitoring details, including method or technique (i.e., visual, volumetric, surface inspection), 

frequency, sample size, data collection, and timing of inspections to ensure timely detection of 

aging effects. In general, the information in this element describes the “when,” “where,” and 

“how” of the AMP (i.e., the specific aspects of the activities to collect data as part of the 

inspection or monitoring activities).

The reviewer should ensure that the applicant has provided sufficient details on the following 

aspects of the inspection or monitoring activities:

• Method or technique: Consistent with quality assurance requirements, the method 

should be demonstrated to be capable of evaluating the condition of the SSC against 

the acceptance criteria for the specific aging mechanism or effect being monitored or 

inspected (as defined in AMP Element 6). For example, the applicant should provide 

a valid technical basis that a particular visual inspection method or instrument has 

sufficient resolution to identify a specific crack or defect dimension. Inspections should 

utilize consensus codes and standards, as applicable.

• Frequency: The reviewer should ensure that the proposed intervals for inspection or 

monitoring are consistent with applicable site-specific, design-specific, or industrywide 

operating experience. Inspections should have sufficient frequency to ensure that 

intended functions will be maintained until the next scheduled inspection.

• Sample size and selection of SSCs for inspection: For a limited sample size, the 

applicant should identify and justify the number of SSCs to be evaluated per inspection, 

including the extent of the inspection for each SSC (e.g., all accessible areas of 

five concrete overpacks in service), and criteria for selection of the specific SSCs for 

inspection based on parameters that may contribute to the operable aging mechanisms 
35

and effects. Consideration should also be given to event-driven fabrication or 

operational issues that may contribute to degradation when selecting SSCs for 

inspection (e.g., welding repairs, occurrence of natural or man-induced events, exposure 

to potentially corrosive environments before the storage term, duration of time between 

fabrication of an SSC and the start of the storage term). The reviewer should ensure the 

applicant has justified using a limited sample size (e.g., one cask per pad) with a 

technical basis, which should include applicable site-specific, design-specific and 

industrywide operating experience. The application should also define the areas that 

have been determined to be inaccessible or below-grade and propose how the condition 

of the inaccessible SSCs will be assessed. The reviewer should ensure that the scope 

of each inspection is properly defined for both accessible and inaccessible (including 

below-grade) areas.

• Data collection: The application should reference any specific methods to be used for 

data acquisition and documentation, including any applicable consensus codes and 

standards. For example, the application may reference field evaluation guides for 

evaluating and documenting cracks in concrete (e.g., ACI 224.1R, ACI 201.1R).

• Timing of inspections: If pre-application inspections were not performed for each of the 

sites covered by or using the subject license or CoC, the reviewer should ensure that 

AMPs include baseline inspections at each site upon entering the period of extended 

operation. Baseline inspections, or the first inspections conducted within the AMPs, 

should assess the condition of SSCs to confirm the results of the pre-application 

inspections that were conducted at other sites or to verify the technical justification 

provided in the application when pre-application inspections were not performed. The 

reviewer should also consider any specific information on the proposed inspection 

schedule (e.g., time of the year) to support the effective detection of aging effects before 

a loss of intended function.

3.6.1.5 Monitoring and Trending

Monitoring and trending should provide for an evaluation of the extent of the effects of aging 

and the need for timely corrective or mitigative actions. This element describes how the data 

collected will be evaluated. This includes an evaluation of the results against the acceptance 

criteria and an evaluation regarding the rate of degradation to ensure that the timing of the next 

scheduled inspection will occur before a loss of intended function. For most cases, this element 

should have a baseline established before or at the beginning of the period of extended 

operation. The reviewer should determine if a baseline inspection is necessary to establish the 

parameters to be monitored and the trending analysis and, if so, whether the proposed baseline 

inspection is adequate.

Although the licensee may have the flexibility to inspect different SSCs or SSC subcomponents 

during subsequent inspections, the selection of different components for the inspections defined 

in the AMP relative to the components inspected in the baseline inspection may present issues 

for purposes of monitoring and trending. If subcomponents from different systems will be 

inspected over the period of extended operation, the monitoring and trending element should 

address how a given operable degradation mode will be adequately trended.
36

3.6.1.6 Acceptance Criteria

The acceptance criteria, against which the need for corrective action will be evaluated, should 

ensure that the SSC intended functions and the approved design bases are maintained during 

the period of extended operation. The proposed acceptance criteria should be appropriately 

justified.

The acceptance criteria could be specific numerical values or could consist of a discussion of 

the process for calculating specific numerical values of conditional acceptance criteria to ensure 

that the design bases are maintained. The reviewer should ensure that the acceptance criteria:

• Include a quantitative basis (justifiable by operating experience, engineering analysis, 

consensus codes/standards).

• Avoid use of non-quantifiable phrases (e.g., significant, moderate, minor, little, slight, 

few; see Appendix A).

• Are achievable and actionable.

The acceptance criteria may be taken directly from the design bases information included in 

either the final safety analysis report (FSAR) or technical specifications. The acceptance criteria 

also may be established by methods provided in NRC-approved topical reports or appropriate 

codes and standards.

3.6.1.7 Corrective Actions

Corrective actions are the measures to be taken when the acceptance criteria are not met. 

Timely corrective actions, including root cause determination and prevention of recurrence for 

significant conditions adverse to quality, are critical for maintaining the intended functions of the 

SSCs during the initial storage period as well as the period of extended operation.

Corrective actions should be described in adequate detail or referenced to source documents. 

An applicant may reference the use of a Corrective Action Program (CAP) approved under

10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel 

Reprocessing Plants,” or 10 CFR Part 72, Subpart G, “Quality Assurance,” to capture and 

address aging effects identified in the period of extended operation. At a minimum, all 

conditions that do not meet the AMP acceptance criteria should be entered into the CAP. The 

QA Program ensures that corrective actions are completed within the specific or general 

licensee’s CAP and includes provisions (as applicable) to:

• Perform functionality assessments.

• Perform apparent cause evaluations and root cause evaluations.

• Address the extent of condition.

• Determine actions to prevent recurrence for significant conditions adverse to quality.

• Justify non-repairs.
37

• Ensure corrective actions are adequately and effectively performed and do not have an 

adverse effect (aging-related or otherwise) on the subject component or other SSCs.

• Trend conditions.

The CAP should be able to respond to and adequately address any ISFSI or DSS aging issues. 

Also, the CAP’s response to addressing any ISFSI or DSS aging issues should include any 

specific corrective actions specified in the license or CoC renewal application.

In some cases, the reviewer may determine the need for specific corrective actions, rather than 

referring only to the use of the CAP. For example, when very limited information exists on the 

condition of an SSC or the applicability of an aging effect at the time of the application, the 

application should include specific follow-up activities when AMP acceptance criteria are not 

met. Thus, the reviewer should review the corrective actions element with consideration of the 

safety-significance of the SSC and the ISFSI site-specific, DSS design-specific, and 

industrywide operating experience to ensure that the proposed corrective actions are 

adequate and effective.

3.6.1.8 Confirmation Process

This element of the program is intended to verify that preventive actions are adequate and 

that appropriate corrective actions have been completed and are effective. The confirmation 

process is commensurate with the specific or general licensee QA Program approved under

10 CFR Part 72, Subpart G, or 10 CFR Part 50, Appendix B. The QA Program ensures that the 

confirmation process includes provisions to preclude repetition of significant conditions adverse 

to quality. The reviewer should ensure the confirmation process describes or references 

procedures to:

• determine follow-up actions to verify effective implementation of corrective actions

• monitor for adverse trends due to recurring or repetitive findings or observations

The reviewer should be aware that the effectiveness of prevention and mitigation programs 

should be verified periodically. For example, in managing corrosion of carbon steel 

components, a mitigation program (coating) may be used to minimize susceptibility to corrosion. 

However, it also may be necessary to have a condition monitoring program (visual or other 

types of inspections) to verify that corrosion is being prevented or controlled to prevent a loss 

of intended function.

3.6.1.9 Administrative Controls

Administrative controls provide a formal review and approval process. Thus, any aging 

management programs must be administratively controlled and included in the FSAR 

supplement. The administrative controls are in accordance with the specific or general licensee 

QA Program approved under 10 CFR Part 72, Subpart G or 10 CFR Part 50, Appendix B, 

respectively. The QA Program ensures that the administrative controls include provisions, 

such as those that define:

• instrument calibration and maintenance

• inspector requirements

• record retention requirements

• document control
38

3.6.1.10 Operating Experience

The reviewer should verify that the operating experience element of the program supports a 

determination that the effects of aging will be adequately managed so that the SSC intended 

functions will be maintained during the period of extended operation. Operating experience is 

useful in providing justification for the effectiveness of each AMP program element and critical 

feedback for enhancement. The reviewer should verify that any degradation in the referenced 

operating experience has been clearly identified as either age-related or event-driven, with 

proper justification for that assessment.

The reviewer should verify that the AMP references and evaluates applicable operating 

experience, including, but not necessarily limited to:

• internal and industrywide condition reports

• relevant international and non-nuclear operating experience

• pre-application inspection results (see below)

• licensee event reports

• vendor-issued safety bulletins

• NRC Generic Communications

• updated consensus codes, standards, or guides

• applicable industry-initiatives (e.g., Department of Energy or Electric Power Research 

Institute sponsored inspections)

The reviewer should consider operating experience of existing programs, including past 

corrective actions resulting in program enhancements or additional programs. A past failure 

would provide feedback from operating experience and should have resulted in appropriate 

program enhancements or new programs. This information can demonstrate where an existing 

program has been implemented correctly and where it has shortcomings in detecting 

degradation in a timely manner. This information should provide objective evidence to support 

or refute the conclusion that the effects of aging will be managed adequately so that the SSC 

intended functions will be maintained during the period of extended operation.

Pre-Application Inspections

Because inspections of DSSs and ISFSI SSCs are not typically conducted during the initial 

storage period, or are narrow in scope, the reviewer may have limited information regarding the 

extent and rate of operable degradation mechanisms for the specific design, environment, and 

operational parameters. In this case, the reviewer should ensure that the scope, methods, 

frequencies and acceptance criteria for monitoring and inspection activities in the aging 

management programs are sufficiently conservative, such that a loss of SSC-intended functions

does not occur during the period of extended operation. Pre-application inspections are one 

means by which an applicant can provide operating experience to justify the specific information 

in the 10 elements of the proposed aging management programs. In addition, pre-application 

inspections may provide data to support a TLAA or other analysis that justifies that an aging 
39

effect will not challenge a SSC’s ability to perform its function(s) (see Section 3.5). The staff 

considers the results of a pre-application inspection, in conjunction with other relevant operating 

experience in its review of the renewal application.

Section 3.4.1.2 discusses the acceptable characteristics of a pre-application inspection, 

including the selection of bounding systems, the use of surrogate inspections, and ensuring 

that the inspection scope, methods, and acceptance criteria align with the guidance for the AMP 

elements described in Section 3.6.1 of this document. Section 3.4.1.2 also discusses the use of 

baseline inspections in an AMP upon entering the period of extended operation to assess the 

condition of SSCs at sites that were not the subject of a pre-application inspection. 

Furthermore, when pre-application inspections are not performed, the reviewer should consider 

whether uncertainties in the degradation of SSCs warrant the addition of conditions to the 

license or CoC to require the baseline and other specific inspections to ensure the AMPs are 

adequate, including for all potential user sites in the case of a renewed CoC (see Section 1.4.7).

Learning AMPs

The reviewer should ensure that the application includes provisions to conduct future reviews of 

site-specific, design-specific, and industrywide operating experience, including relevant 

international and non-nuclear operating experience, to confirm the effectiveness of the AMPs 

or identify a need to enhance or modify an AMP. The reviewer should verify that the applicant: 

(1) references a specific system to be used to obtain, aggregate, and enter site-specific, 

design-specific, and industrywide operating experience (e.g., Institute of Nuclear Power 

Operations database), and (2) has discussed how it intends to provide timely reporting of 

operating experience to this system.

The commitment to future reviews should ensure that, as knowledge and data become available 

from new analyses, experiments, and operating experience, the licensees and CoC holders will

revise the existing AMPs (or pertinent procedures for AMP implementation) as necessary to 

address any lessons learned identified during the review of operating experience.

If an applicant follows this approach, the reviewer should ensure that the description of the 

periodic assessments includes specific performance criteria (e.g., program-specific performance 

indicators for each of the 10 AMP elements) and proposed actions based on the assessment 

findings. The reviewer should also ensure that the timing of the assessments appropriately 

considers the rate of aging degradation and the anticipated availability of data from industry 

initiatives. The reviewer should consider the frequency, acceptance criteria, and proposed 

corrective actions for these assessments for the requisite finding of reasonable assurance.

Nuclear Energy Institute (NEI) 14-03, “Format, Content, and Implementation Guidance for Dry 

Cask Storage Operations-Based Aging Management,” Revision 1, provides a proposed 

framework for learning AMPs through the use of “tollgates.” NEI 14-03 defines “tollgates” as 

periodic points within the period of extended operation when licensees would be required to 

evaluate aggregate feedback and perform and document a safety assessment that confirms 

the safe storage of spent fuel. Tollgates are described as an additional set of in-service 

assessments beyond the normal continual assessment of operating experience, research, 

monitoring, and inspections on DSS component and ISFSI SSC performance that is part of 

normal ISFSI operations for licensees during the initial storage period as well as the period of 

extended operation. The reviewer should be aware that an applicant may reference the use of 

“tollgates” in the renewal application. The reviewer should (1) assess the applicant’s proposed 

periodic assessments of operating experience and other relevant data, and (2) make a 
40

determination regarding the ability of these assessments to ensure the continued effectiveness 

of AMPs. The reviewer should ensure that tollgates determined to be necessary to demonstrate 

the continued effectiveness of the AMPs are included in the application.

NEI 14-03, Revision 1, also describes a framework for the aggregation and dissemination 

of operating experience across the industry through the use of an aging-related operating 

experience “clearinghouse,” titled the ISFSI Aging Management Institute of Nuclear Power 

Operations Database (ISFSI AMID). Whether the applicant references the ISFSI AMID 

described in NEI 14-03 or proposes an alternative means to seek out operating experience, the 

reviewer should ensure that the application describes how industrywide operating experience 

and results of industry initiatives will be accessed, utilized, and assessed to ensure that AMPs 

are modified as appropriate.

At the time of publication, the staff was continuing its review of NEI 14-03, Revision 1, for 

proposed NRC endorsement, as discussed in the Introduction.

The NRC will inspect licensees’ implementation of AMPs in the period of extended operation, 

including any licensee actions taken as part of the “learning” aspect of AMPs. The NRC will 

inspect licensees’ periodic assessments of AMP effectiveness and any adjustments licensees 

have made to AMPs to respond to operating experience and ensure AMPs are effective for

addressing aging effects in the period of extended operation.

3.6.2 Commencement of AMP(s) for CoC Renewals

An AMP for a renewed CoC commences at the end of the initial storage period for each loaded 

DSS (see Appendix F for discussion of storage terms, including an explanation of when storage 

begins). Activities in an AMP may start at different timeframes (e.g., an AMP can be 

implemented before the period of extended operation to capture baseline data and AMPs may 

have different frequencies of implementation). Additional considerations for CoC renewals and 

general licensee implementation of AMPs are provided in Appendix E.

3.6.3 Implementation of AMP(s)

Generally, licensees should develop the infrastructure for AMP implementation (e.g., procure 

equipment or contracts, train personnel, or update, revise, or develop procedures for 

implementing AMP activities) before entering the period of extended operation. However, 

this may not be possible when the initial storage term ends shortly after the license or CoC is 

renewed or if a license or CoC is in the period of timely renewal (Section 1.4.5). In such cases, 

the development of the infrastructure for AMP implementation generally should be no later than 

one year from the date the NRC issues a renewed specific license or CoC. The reviewer should 

ensure that the timing of implementation for each AMP is addressed in the application in a clear 

manner and is appropriately justified if it exceeds the above guidance. Additional 

considerations for CoC renewals and general licensee implementation of AMPs are provided in 

Appendix E.

3.6.4 Evaluation Findings

The reviewer prepares the summary statement and evaluation finding based on compliance 

with the regulatory requirements in Section 3.3. The summary statement and evaluation finding 

should be similar in wording to the following example (the finding number is for convenience in 

cross-referencing within the SRP and SER):
41

The NRC staff reviewed the aging management programs provided in the renewal application 

and supplemental documentation. The NRC staff performed its review following the guidance 

provided in NUREG-1927, Revision 1, “Standard Review Plan for Renewal of Specific Licenses 

and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel” and relevant ISGs. 

Based on its review, the NRC staff finds:

F3.4 The applicant has identified programs that provide reasonable assurance that aging 

effects will be managed effectively during the period of extended operation, in accordance with 

10 CFR 72.42(a)(2) or 10 CFR 72.240(c)(3), as applicable.

43

4. CONSOLIDATED REFERENCES

Title 10 of the Code of Federal Regulations (10 CFR) Part 2, “Agency Rules of Practice and 

Procedure,” Washington, DC.

10 CFR Part 20, “Standards for Protection against Radiation,” Washington, DC.

10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” Washington, DC.

10 CFR Part 51, “Environmental Protection Regulations for Domestic Licensing and Related 

Regulatory Functions,” Washington, DC.

10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” 

Washington, DC.

10 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,” Washington, 

DC.

American Concrete Institute (ACI) 349.3R-02, “Evaluation of Existing Nuclear Safety-Related 

Concrete Structures” (Reapp 2010).

ACI 318-05, “Building Code Requirements for Structural Concrete and Commentary,” 2005.

ACI 349-06, “Code Requirements for Nuclear Safety-Related Concrete Structures and 

Commentary,” 2007.

ACI 201.1R-8, “Guide for Conducting a Visual Inspection of Concrete in Service,” 2008.

ACI 224.1R-07, “Causes, Evaluation, and Repair of Cracks in Concrete Structures,” 2007.

ACI 562-13, “Code Requirements for Evaluation, Repair, and Rehabilitation of Concrete 

Buildings and Commentary,” 2013.

ACI 364.1R-07, “Guide for Evaluation of Concrete Structures before Rehabilitation,” 2007.

ACI 207.3R-94, “Practices for Evaluation of Concrete in Existing Massive Structures for Service 

Conditions,” 2008.

ACI 506R-05, “Guide to Shotcrete,” 2005.

ACI CT-13, “ACI Concrete Terminology,” 2013.

American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) 11-99 (2000), 

“Guideline for Structural Condition Assessment of Existing Buildings.”

American Society of Mechanical Engineers (ASME), “Boiler and Pressure Vessel Code Section 

XI—Rules for Inservice Inspection of Nuclear Power Plant Components,” New York, NY, 2007.

ASME Code Section XI, Subsection IWL, “Requirements for Class CC Concrete Components of 

Light-Water-Cooled Plants,” 2013.
44

Bare, W.C, L.D. Torgerson, “Dry Cask Storage Characterization Project—Phase 1: 

CASTOR V/21 Cask Opening and Examination,” NUREG/CR-6745, Idaho National Engineering 

and Environmental Lab, Idaho Falls, ID, 2001, Agencywide Documents Access and 

Management System (ADAMS) Accession No. ML013020363.

Billone, M.C., T.A. Burtseva, and R.E. Einziger, “Ductile-to-Brittle Transition Temperature for 

High-Burnup Cladding Alloys Exposed to Simulated Drying-Storage Conditions,” Journal of 

Nuclear Materials, Volume 433, Issues 1–3, February 2013, pages 431–448.

Cumblidge, S.E., M.T. Anderson, S.R. Doctor, “An Assessment of Visual Testing,”

NUREG/CR-6860, Pacific Northwest National Laboratory, Richland, WA, 2004,

ADAMS Accession No. ML043630040.

Cumblidge, S.E, et al., “A Study of Remote Visual Methods to Detect Cracking in Reactor 

Components,” NUREG/CR-6943, Pacific Northwest National Laboratory, Richland, WA, 2007,

ADAMS Accession No. ML073110060.

Daum, R.S., et al., “Radial-Hydride Embrittlement of High-Burnup Zircaloy-4 Fuel 

Cladding”, Journal of Nuclear Science and Technology, Vol. 43, No. 9, 2006, p. 1054.

Dominion Resources, Inc., “North Anna ISFSI TN-32 High Burnup Dry Storage Research 

Project,” May 13, 2015, ADAMS Accession No. ML15133A082.

Einziger, R.E., et al., “Examination of Spent PWR Fuel Rods after 15 Years in Dry Storage,”

NUREG/CR-6831, Argonne National Laboratory, Argonne, IL., 2003, ADAMS Accession 

No. ML032731021.

Electric Power Research Institute (EPRI)/U.S. Department of Energy (DOE), HBU Dry Storage 

Cask Research and Development Project Final Test Plan, February 27, 2014, DOE Contract 

No. DE-NE-0000593. Accessible at http://www.energy.gov/sites/prod/files/2014/03/f8/

HBUDry%20StrgeCaskRDfinalDemoTestPlanRev9.pdf.

Gorman, J., K. Fuhr, J. Broussard, “Literature Review of Environmental Conditions and 

Chloride-Induced Degradation Relevant to Stainless Steel Canisters in Dry Cask Storage 

Systems,” EPRI-3002002528, Palo Alto, CA, 2014.

Hayashibara, H., et al., “Effect of Temperature and Humidity on Atmospheric Stress Corrosion 

Cracking of 304 Stainless Steel,” NACE International, Corrosion 2008, paper 08492, Houston, 

TX, 2008.

He, X., et al., “Assessment of Stress Corrosion Cracking Susceptibility for Austenitic Stainless 

Steels Exposed to Atmospheric Chloride and Non-Chloride Salts,” NUREG/CR-7170, 

U.S. Nuclear Regulatory Commission (NRC), 2014, ADAMS Accession No. ML14051A417.

Kamimura, K., “Integrity Criteria of Spent Fuel for Dry Storage in Japan,” Proceedings of

International Seminar on Interim Storage of Spent Fuel, ISSF 2010, Tokyo Nov 2010, 

page VI3-1.

Kosaki, A., “Evaluation Method of Corrosion Lifetime of Conventional Stainless Steel Canister 

under Oceanic Air Environment,” Nuclear Engineering and Design, Vol. 238, 2008, 

pp. 1233-1240.
45

McConnell, J.W. Jr., A.L. Ayers, Jr., M.J. Tyacke, “Classification of Transportation Packaging 

and Dry Spent Fuel Storage System Components According to Importance to Safety,” 

NUREG/CR-6407, NRC, 1996. Accessible at 

http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6407/cr6407.pdf.

Nuclear Energy Institute (NEI), “Guidance for Operations-Based Aging Management for 

Dry Cask Storage,” NEI 14-03, Rev. 0, 2014, ADAMS Accession No. ML14266A225.

NEI, “Format, Content and Implementation Guidance for Dry Cask Storage Operations-Based 

Aging Management for Dry Cask Storage,” NEI 14-03, Rev. 1, 2015, ADAMS Accession 

No. ML15272A329.

NRC,

3 “Standard Review Plan for Spent Fuel Dry Storage Facilities,” NUREG-1567, Rev. 0,

Washington, DC, 2000, ADAMS Accession No. ML003686776.

NRC, “10 CFR Part 72, License and Certificate of Compliance Terms, Final Rule,” 

Federal Register, Vol. 76, No. 32, February 16, 2011, pp. 8872–8892 (76 FR 8872).

NRC, “Cladding Considerations for the Transportation and Storage of Spent Fuel,” ISG-11, 

Rev. 3, Washington, DC, 2003, ADAMS Accession No. ML033230335.

NRC, Regulatory Issue Summary (RIS) 2004-20, “Lessons Learned from Review of 

10 CFR Parts 71 and 72 Applications,” December 16, 2004, ADAMS Accession 

No. ML043510074.

NRC, “Standard Review Plan for Spent Fuel Dry Storage Systems at a General License 

Facility,” NUREG-1536, Rev. 1, Washington, DC, 2010a, ADAMS Accession No. ML101040620.

NRC, “Fuel Retrievability,” ISG-2, Rev. 1, Washington, DC, 2010b, ADAMS Accession 

No. ML100550861.

NRC, “Generic Aging Lessons Learned (GALL) Report,” NUREG-1801, Rev. 2, Washington, 

DC, 2010c, ADAMS Accession No. ML103490041.

NRC, “Standard Review Plan for Renewal of Spent Fuel Dry Cask Storage System 

Licenses and Certificates of Compliance,” NUREG-1927, Rev. 0, Washington, DC, 2011,

ADAMS Accession No. ML111020115.

NRC, “Clarifying the Relationship Between 10 CFR 72.212 and 10 CFR 72.48 Evaluations,” 

RIS 2012-05, 2012a, ADAMS Accession No. ML113050537.

NRC, “Potential Chloride-Induced Stress Corrosion Cracking of Austenitic Stainless Steel and 

Maintenance of Dry Cask Storage System Canisters,” Information Notice 2012-20, 2012b,

ADAMS Accession No. ML12319A440.

NRC, “The Use of a Demonstration Program as a Surveillance Tool for Confirmation of Integrity 

for Continued Storage of High Burnup Fuel Beyond 20 Years,” ISG-24, Rev. 0, Washington, DC, 

2014, ADAMS Accession No. ML14058B166.

 3 Note that references to NRC NUREG reports may be superseded by later versions. References to ISGs 

may be superseded by incorporation into NUREG reports.
46

NRC, Letter from M. Lombard (NRC) to R. McCullum (NEI), Response to September 23, 2014, 

Nuclear Energy Institute Submittal: NEI 14-03, “Guidance for Operations-Based Aging 

Management for Dry Cask Storage,” Rev. 0, dated January 21, 2015, ADAMS Accession 

No. ML15013A201.

NRC, Request for public comments on draft NUREG, “Standard Review Plan for Renewal 

of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel,” 

Federal Register, Vol. 80, No. 129, July 7, 2015, pp. 38780–38782 (80 FR 38780).

Selby, G. “BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals 

Examination Guidelines,” EPRI 1011689, TR-105696-R8 (BWRVIP-03) Rev. 8, Palo Alto, 

CA, 2005.

Sheaffer, M.K., et al., “Engineering Drawings for 10 CFR Part 71 Package Approvals,”

NUREG/CR-5502, Washington, DC, 1998. Accessible at https://www.orau.org/ptp/pdf/5502.pdf.

Tokiwai, M., H. Kimura, H. Kusanagi, “The Amount of Chlorine Contamination for Prevention of 

Stress Corrosion Cracking in Sensitized Type 304 Stainless Steel,” Corrosion Science, Vol. 25,

Issue 8–9, 1985, pp. 837–844.

Waldrop, K., et al., “Calvert Cliffs Stainless Steel Dry Storage Canister Inspection,” 

EPRI-1025209, EPRI, Palo Alto, CA, 2014.
47

5. GLOSSARY

Accident condition: The extreme level of an event or condition, which has a specified 

resistance, limit of response, and requirement for a given level of continuing capability, which 

exceeds off-normal events or conditions. Accident conditions include both design-basis 

accidents and conditions caused by natural and manmade phenomena.

Aging effect: The manifestation of an aging mechanism (e.g., cracking, loss of fracture 

toughness, loss of material).

Aging management activity (AMA): An application of either the aging management program 

(AMP) or time-limited aging analyses (TLAAs) to provide reasonable assurance that the 

intended functions of structures, systems, and components (SSCs) of independent spent fuel 

storage installations (ISFSIs) and dry storage systems (DSSs) are maintained during the period 

of extended operation.

Aging management program (AMP): A program for addressing aging effects that may include 

prevention, mitigation, condition monitoring, and performance monitoring. See Title 10 of the 

Code of Federal Regulations (10 CFR) 72.3, “Definitions.”

Aging management review (AMR): An assessment conducted by the licensee or certificate of 

compliance (CoC) holder that addresses aging mechanisms and effects that could adversely 

affect the ability of SSCs from performing their intended functions during the period of 

extended operation.

Aging mechanism: The degradation process for a given material and environment which 

results in an aging effect (e.g., freeze-thaw degradation, neutron irradiation, erosion).

Amendment of a license or CoC: An application for amendment of a license or a CoC must be 

submitted whenever a holder of a specific license or CoC desires to amend the license or CoC 

(including a change to the license or CoC conditions). The application must fully describe the 

changes desired and the reasons for such changes, and following as far as applicable the form 

prescribed for original applications. See 10 CFR 72.56, “Application for Amendment of

License,” and 10 CFR 72.244, “Application for Amendment of a Certificate of Compliance”.

Baseline inspection: The first inspection of an AMP to assess the condition of SSCs to either: 

(1) confirm that the results of pre-application inspections conducted at other sites are bounding 

of the subject site, or (2) verify the adequacy of the AMPs and the conclusions of the TLAAs 

when pre-application inspections were not performed.

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

through fission, usually expressed in GWd/MTU (gigawatt days per metric ton uranium).

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 spent nuclear fuel (SNF) 

required by the applicable regulations.
48

Canister (in a dry storage system for SNF): 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. See NUREG-1536, “Standard 

Review Plan for Spent Fuel Dry Storage Systems at a General License Facility.”

Certificate of compliance (CoC) (for a dry storage system for SNF): The certificate issued 

by the NRC that approves the design of a spent fuel storage cask in accordance with the 

provisions of 10 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,” Subpart L, “Approval of Spent Fuel Storage Casks.” See 10 CFR 72.3.

Certificate of compliance holder (CoC holder): A person who has been issued a CoC by the 

U.S. Nuclear Regulatory Commission (NRC) for a spent fuel storage cask design under 

10 CFR Part 72. See 10 CFR 72.3.

Certificate of compliance user (CoC user): The general licensee that has loaded, or plans to 

load, a dry storage system (DSS) in accordance with a CoC issued under 10 CFR Part 72.

Confinement (in a dry storage system for spent nuclear fuel): The ability to limit or prevent 

the release of radioactive substances into the environment.

Confinement systems: Those systems, including ventilation, that act as barriers between 

areas containing radioactive substances and the environment. See 10 CFR 72.3.

Controlled area: The area immediately surrounding an ISFSI for which the licensee exercises 

authority over its use and within which it performs ISFSI operations. See 10 CFR 72.3.

Criticality: The condition wherein a system or medium is capable of sustaining a nuclear 

chain reaction.

Degradation: Any change in the properties of a material that adversely affects the performance 

of that material; adverse alteration. See NUREG-1536.

Design bases:

4

 Information that identifies the specific function(s) to be performed by SSCs 

(both important-to-safety and not important-to-safety) of a facility or of a spent fuel storage cask 

and the specific values or ranges of values chosen for controlling parameters as reference 

bounds for design. These values may be (1) restraints, derived from generally accepted 

“state-of-the-art” practices for achieving functional goals, or (2) requirements, derived from 

analysis (based on calculation, experiments, or both) of the effects of a postulated event under 

which SSCs must meet their functional goals. See 10 CFR 72.3.

Dry storage: The storage of spent nuclear fuel in a DSS, which typically involves drying the 

DSS cavity and backfilling with an inert gas.

 4 The NRC has removed references to either “current licensing basis” (CLB) or “licensing basis”, and replaced 

them with “design bases” in this guidance revision. Neither “current licensing basis” nor “licensing basis” is 

defined in 10 CFR Part 72. In addition, in the Statement of Considerations for the 2011 Part 72 rulemaking 

change (NRC, 2011), the NRC stated “The NRC does not believe that it is appropriate for the CLB to be 

applied to cask CoC renewals, which are generic.”
49

Dry storage system (DSS): A system that typically uses a cask or canister in an overpack as a 

component in which to store spent nuclear fuel in a dry environment. A DSS provides 

confinement, radiological shielding, sub-criticality control, structural support, and passive 

cooling of its spent nuclear fuel during normal, off-normal, and accident conditions.

General license: Authorizes the storage of spent fuel in an ISFSI at power reactor sites to 

persons (i.e., general licensee) authorized to possess or operate nuclear power reactors under 

10 CFR Part 50 (“Domestic Licensing of Production and Utilization Facilities”) or Part 52 

(“Licenses, Certifications, and Approvals for Nuclear Power Plants”). The general license is 

limited to (1) that spent fuel which the general licensee is authorized to possess at the site 

under the specific Part 50 or Part 52 license for the site, and (2) storage of spent fuel in casks 

approved under the provisions of 10 CFR Part 72, Subpart L. See 10 CFR 72.210 (“General 

License Issued”) and 72.212(a)(1)–(2).

High burnup (HBU) fuel: Spent nuclear fuel with burnups generally exceeding 45 GWd/MTU.

Horizontal storage module: A reinforced, heavy-walled concrete structure designed to store dry 

spent fuel canisters in a horizontal position at an independent spent fuel storage installation. 

The horizontal storage module provides physical protection of canisters and radiological 

shielding, while allowing passive cooling.

Important to safety (ITS): See structures, systems, and components (SSCs) important to safety.

Independent spent fuel storage installation (ISFSI): A complex designed and constructed for 

the interim storage of spent nuclear fuel, solid reactor-related greater-than-Class-C (GTCC)

waste, and other radioactive materials associated with spent fuel and reactor-related GTCC 

waste storage. See 10 CFR 72.3.

Inspection: The examination of an SSC, using a nondestructive testing technique, to determine 

its current condition and if there is any damage, defect, or degradation that could have an 

adverse effect on the function of that SSC.

Intended function: A design-bases function defined as either (1) important to safety or (2) failure

of which could impact a safety function.

Interim staff guidance (ISG): Supplemental information that clarifies important aspects of 

regulatory requirements. An ISG provides review guidance to NRC staff in a timely manner until 

standard review plans are revised accordingly. See NUREG-1536.

Monitored retrievable storage installation (MRS): A complex designed, constructed, and 

operated by the U.S. Department of Energy for the receipt, transfer, handling, packaging, 

possession, safeguarding, and storage of spent nuclear fuel aged for at least 1 year, solidified 

high-level radioactive waste resulting from civilian nuclear activities, and solid reactor-related 

GTCC waste, pending shipment to a HLW repository or other disposal. See 10 CFR 72.3.
50

Monitoring: Data collection (from activities performed in either the initial storage period or the 

period of extended operation) to determine the status of a DSS, ISFSI, or both, and to verify the 

continued efficacy of the system, on the basis of measurements of specified parameters, 

including temperature, direct radiation, radioactive effluents, functionality, and characteristics 

of components of the system. Monitoring could thus be described as those activities that 

periodically or continuously monitor performance as an indirect indicator of degradation 

(e.g., monitoring groundwater chemistry) or monitor the effectiveness of preventive measures. 

With respect to direct radiation and radioactive effluents, according to 10 CFR 20.1003,

“Definitions,” 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. See NUREG-1536.

Normal events or conditions: The maximum level of an event or condition expected to routinely 

occur. Events and conditions that exceed the levels associated with “normal” are considered to 

be, and to have the response allowed for, “off-normal” or “accident-level” events and conditions. 

See NUREG-1567, “Standard Review Plan for Spent Fuel Dry Storage Facilities.”

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 (once per calendar year) 

and for which there is a corresponding maximum specified resistance, specified limit of 

response, or requirement for a specified level of continuing capability. Off-normal is considered 

to include “anticipated occurrences” as used in 10 CFR Part 72. See NUREG-1536.

Overpack: A heavy-walled concrete, metal, or combined concrete and metal structure designed 

to store spent fuel canisters at an ISFSI. The overpack provides physical protection of canisters 

and radiological shielding, while allowing passive cooling.

Pre-application inspection: An inspection performed at the discretion of the licensee or CoC 

holder before submittal of the renewal application to provide operating experience to support 

the aging management review, proposed AMP activities, or evaluation of TLAAs.

Radiation shielding: ISFSI and DSS SSCs that are designed so that dry storage operations 

at an ISFSI meet the requirements of 10 CFR 72.126(a)(6) and 10 CFR 72.128(a)(2) and the 

requirements of 10 CFR 72.104(a) and 10 CFR 72.106(b), when both direct radiation and 

radioactive effluents are considered.

Renewal of a license or CoC: A certificate holder may apply for renewal of the design of a spent 

fuel storage cask for a term not to exceed 40 years. In the event that the certificate holder does 

not apply for a cask design renewal, any licensee using a spent fuel storage cask, a 

representative of the licensee, or another certificate holder may apply for a renewal of that 

cask design for a term not to exceed 40 years. See 10 CFR 72.240, “Conditions for Spent 

Fuel Storage Cask Renewal.” Specific licenses may be renewed by the Commission at the 

expiration of the license term upon application by the licensee for a period not to exceed 

40 years. See 10 CFR 72.42, “Duration of License; Renewal.”

Retrievability: 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. 

See 10 CFR 72.122(l). ISG-2 provides guidance on the fuel retrievability, including ready 

retrieval.
51

Safety analysis report (SAR): The document that a CoC holder, specific licensee, an applicant 

for a CoC, or an applicant for a specific license supplies to the NRC for evaluation. For 

specific-license renewals, the SAR must contain information required in 10 CFR 72.24, 

“Contents of Application; Technical Information.” For CoC renewals, the SAR must meet the 

requirements of 10 CFR 72.240(b). The SAR provides references and drawings of the DSS, 

ISFSI, or both; details of construction; materials; and standards to which the SSC has been 

designed or fabricated. For clarification, SAR is a general term; while FSAR indicates the 

document that is submitted within 90 days after the issuance of the license or CoC that is based 

on the SAR in the license or CoC application and reflects any changes or applicant 

commitments developed during the license or CoC approval and/or hearing process. Both 

FSAR and updated final safety analysis report (UFSAR) are terms that are used to indicate the 

FSAR update that is required every 2 years. A specific licensee or CoC holder shall update the 

FSAR in accordance with 10 CFR 72.70 (“Safety Analysis Report Updating”) or 10 CFR 72.248, 

(“Safety Analysis Report Updating”) respectively.

Safety evaluation report (SER): The document that the NRC publishes at the completion of a 

licensing or certification review. The SER contains all of the NRC staff findings and conclusions 

from the licensing or certification review.

Safety function: A function defined as ITS. The ITS functions that structures, systems, and 

components are designed to maintain include:

• structural integrity

• content temperature control (i.e., heat-removal capability)

• radiation shielding

• confinement

• sub-criticality control

• retrievability

See NUREG-1536.

Service conditions: Conditions (e.g., time of service, temperatures, environmental 

conditions, radiation, and loading) that a component experiences during storage.

Specific license: A license for the receipt, handling, storage, and transfer of spent fuel, 

high-level radioactive waste, or reactor-related GTCC waste that is issued to a named person 

(i.e., specific licensee) on an application filed under regulations in 10 CFR Part 72, Subpart B, 

“License Application, Form, and Contents.”

Spent fuel storage cask or cask: All the components and systems associated with the container 

in which spent fuel, or other radioactive materials associated with spent fuel, is stored at an 

ISFSI. See 10 CFR 72.3.

Spent nuclear fuel or spent fuel: Nuclear fuel that has been withdrawn from a nuclear reactor 

after irradiation, has undergone at least a 1-year decay process 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. See 10 CFR 72.3.
52

Structures, systems, and components (SSCs) important to safety: See 10 CFR 72.3. Those 

features of the ISFSI and spent fuel storage cask whose functions are at least one of the 

following:

• to maintain the conditions required to safely store spent fuel, high-level radioactive 

waste, or reactor-related GTCC waste

• to prevent damage to the spent fuel, the high-level radioactive waste, or reactor-related 

GTCC waste container during handling and storage

• to provide reasonable assurance that spent fuel, high-level radioactive waste, or 

reactor-related GTCC waste can be received, handled, packaged, stored, and retrieved 

without undue risk to the health and safety of the public

Time-limited aging analysis (TLAA): See 10 CFR 72.3. A licensee or CoC holder calculation or 

analysis that has all of the following attributes:

• involves SSCs important to safety within the scope of license or CoC renewal

• considers the effects of aging

• involves time-limited assumptions defined by the current operating term, for example, 

40 years

• was determined to be relevant by the licensee or CoC holder in making a 

safety determination

• involves conclusions or provides the basis for conclusions related to the capability of the 

SSCs to perform their intended safety functions

• is contained or incorporated by reference in the design bases

Transfer cask: A shielded SSC used to transfer the fuel canister between the spent fuel 

handling area and the overpack or storage module.
APPENDIX A

NON-QUANTIFIABLE TERMS

A-1

Appendix A

Non-Quantifiable Terms

It is preferred that renewal applications use quantifiable terms and quantitative information, 

where it exists. However, the following non-quantifiable terms, as well as others, may appear 

in the renewal application, safety analysis report (SAR), and updates to final SARs:

• large

• small

• slight

• slightly

• significant

• significance

• moderate

• moderately

• low 

• minor

• many

• few

• little

• routine

• some

• major

• undetectable

• visible

• measurable

• unchanged

• changed

• no loss of

Table A-1 may be used as guidance for the terms listed above, for additional consideration, or 

to provide quantitative measures or information.
A-2

Table A-1. Screening Criteria for Non-Quantifiable Terms

Terms Actions

Screened In

The term requires additional consideration if it is 

used for one of the following:

• characterizing an aging effect

(e.g., degradation, cracking, fatigue, corrosion, 

loss of material, change in material properties)

• providing important information about the 

operations, functions, or other characteristics 

of an in-scope SSC

• describing dose, environmental impact, or 

other hazard, such as combustible material

or dust

If the term screens in, one of 

the following must be provided:

• quantitative information, if it 

is available

• additional descriptions

• definition of the meaning 

of the term

(e.g., “insignificant” means 

the function of the SSC is 

not impaired)

Screened Out

The term is considered immaterial to the SAR and 

ISFSI/DSS UFSAR for one of the following 

reasons:

• The term is included in the title of 

reference document.

• The term is included in a quote.

• The term is explained by adjacent 

quantitative information (e.g., small:

less than 20 percent).

• Use of the term is NOT related to any of 

the following:

– in-scope SSCs per AMR results

– aging effect

– dose, environment impact, or other hazard

(e.g., combustible material)

• Use of the term does not provide important 

information. It is merely descriptive and the 

meaning of the statement is not changed if 

the term is deleted (e.g., the word “small” 

could be deleted from the following statement 

without altering the meaning: “Water in the 

grapple ring is drained through a small hole”).

No action
APPENDIX B

EXAMPLES OF AGING MANAGEMENT PROGRAMS

B-1

Appendix B

Examples of Aging Management Programs

Appendix B contains example aging management programs (AMPs) for:

• localized corrosion and stress corrosion cracking of welded stainless steel dry storage 

canisters (Table B-1)

• reinforced concrete structures (Table B-2)

• high burnup (HBU) fuel monitoring and assessment program (Table B-3)

This appendix provides examples of acceptable AMPs for staff reference during review 

of renewal applications.
B-2

Example AMP for Localized Corrosion and Stress Corrosion Cracking of Welded 

Stainless Steel Dry Storage Canisters

Welded stainless steel canisters are used in the majority of the dry storage systems in the 

United States for spent nuclear fuel from commercial power reactors at both specific-licensed

and general-licensed independent spent fuel storage installations (ISFSIs). The welded 

stainless steel canisters are the primary confinement boundary during storage. Although there 

are no known operational occurrences of aging or localized corrosion of welded stainless steel 

canisters, operating experience with nuclear reactors that were located close to an open ocean 

or bay has shown that pitting corrosion, crevice corrosion, and chloride-induced stress corrosion 

cracking (CISCC) can occur in welded stainless steel components as a result of atmospheric 

deposition and deliquescence of chloride-containing salts. Laboratory and natural exposure 

tests suggest that CISCC can occur with sufficient surface chloride concentrations and that, with 

those concentrations of chloride, crack propagation rates can be of engineering significance for 

welded stainless steel canisters during the period of extended operation.

Based on reactor operating experience as well as laboratory and field testing, localized 

corrosion and CISCC are potential aging mechanisms for welded stainless steel canisters.

Environments where chloride-containing salts may be deposited on welded stainless steel 

canisters include coastal locations near salt water and locations that are close to cooling towers 

or roads that are salted. ASME Section XI has formed a Task Group to develop a code case to 

establish the requirements for inservice inspection and acceptance criteria for dry storage 

system canisters. However, the development of a consensus based code case for inservice 

inspection of dry storage system canisters may take several years to complete. To address 

potential aging effects as a result of localized corrosion cracking and stress corrosion cracking 

in the absence of an acceptable code case, the U.S. Nuclear Regulatory Commission (NRC)

has provided an example AMP for welded stainless steel canisters used in dry storage systems 

that relies on guidance from consensus codes and standards for inservice inspection of nuclear 

power plant components. Elements of an NRC staff developed example AMP are provided in 

Table B-1.
B-3

Table B-1. Example AMP for Localized Corrosion and Stress Corrosion Cracking of 

Welded Stainless Steel Dry Storage Canisters

Element Description

1. Scope of Program Inspection of welded stainless steel dry storage canister confinement 

boundary external surfaces for atmospheric deposits, localized corrosion, 

and stress corrosion cracking.

Examinations should be focused on areas with the following attributes:

• canister fabrication welds and weld heat affected zones

• closure welds and weld heat affected zones

• areas of the canister to which temporary supports or attachments were 

attached by welding and subsequently removed

• locations where a crevice is formed on the canister surface

• horizontal (±30°) surfaces where deposit accumulation may accumulate 

at a faster rate compared to vertical surfaces

• canister surfaces that are cold relative to the average surface 

temperature

• canister surfaces with higher amounts of atmospheric deposits

Effort should be made to identify and prioritize examinations of areas on 

canisters that have two or more of the above attributes (e.g., canister 

surface that is cold relative to average surface temperature and also has a 

weld/weld heat affected zone).

2. Preventive Actions None, AMP is for condition monitoring. However, dry storage system 

canister designs may include preventative actions, such as fabrication 

procedures and surface modification methods to impart compressive 

residual stresses on the canister welds and weld heat affected zones to 

reduce the potential for stress corrosion cracking. Preventative actions may 

also include the use of dry storage system canister confinement boundary 

materials that are resistant to localized corrosion and stress corrosion

cracking. For such cases the preventative actions described should be 

supported with an analysis and data demonstrating the preventative actions 

are effective.

3. Parameters 

Monitored or

Inspected

Parameters monitored/inspected should include:

• visual evidence of discontinuities and imperfections, such as localized 

corrosion, including pitting corrosion, crevice corrosion and stress 

corrosion cracking of the canister welds and weld heat affected zones

• size and location of localized corrosion and stress corrosion cracks

• appearance and location of deposits on the canister surfaces
B-4

Element Description

4. Detection of 

Aging Effects

Visual examination of deposits on the canister surfaces and to identify 

corrosion products that may be indicators of localized corrosion and stress 

corrosion cracking in the welds and weld heat affected zones. Visual 

examination instrumentation with demonstrated sizing and depth 

measurement capability may be useful in the determination of the size and 

depth of pits open to the surface. Visual examination may also detect the 

presence of cracks originating from pits. However, the ability to detect 

cracks on clean metal surfaces using visual examination methods is 

dependent on several factors and can be difficult for tight crack opening 

displacements (Cumblidge et al., 2004; 2007). The presence of significant 

corrosion product accumulation may also interfere with the identification of 

stress corrosion cracks using visual examination methods.

Volumetric examination is necessary to characterize stress corrosion 

cracking. Volumetric examination of pits and areas immediately adjacent to 

pits is necessary when pits are within 25 mm (1 inch) of a through thickness 

weld or within 25 mm (1 inch) of an area where an temporary attachment 

was known to be located.

Visual Examination

Pitting and crevice corrosion that is open to the surface can potentially be 

detected by visual testing (ASME Section V, Table A-110). Because of the 

high neutron and gamma radiation fields near the surface of the stainless 

steel dry storage canisters, direct visual examination is not possible. 

Procedures for remote visual examination should be performance 

demonstrated; procedure attributes including equipment resolution, lighting 

requirements, etc., should reference applicable standards, such as ASME 

Section XI, Article IWA-2200 for VT-1 and VT-3 examinations (ASME, 2007)

and BWRVIP-03 (Selby 2005) for EVT-1 examinations.

Volumetric Examination

Additional assessment is necessary for suspected areas of localized 

corrosion or stress corrosion cracking. In these cases, the severity of 

degradation must be assessed including the dimensions of the affected area 

and the depth of penetration with respect to the thickness of the canister. 

For accessible areas where adequate cleaning can be performed, remote 

visual examination meeting the requirements for VT-1 Examination 

(ASME Section XI, IWA-2211) may be used to determine the type of 

degradation present (e.g., pitting corrosion or stress corrosion cracking) and 

the location of degradation. Examinations to characterize the extent and 

severity of localized corrosion or stress corrosion cracking should be 

conducted using surface or volumetric examination methods consistent 

with the requirements of ASME Section XI, IWB-2500 for category B-J 

components (ASME, 2007).
B-5

Element Description

4. Detection of Aging 

Effects (cont’d)

Sample Size

For sites where inspections are necessary, a minimum of one canister at 

each site. Preference should be given to the canisters with the greatest 

susceptibility for localized corrosion or SCC. Factors to be considered 

include older and colder canisters with the greatest potential for the 

accumulation and deliquescence of deposited salts that may promote 

localized corrosion and stress corrosion cracking, types of systems used 

at site, canister location with respect to potential sources of atmospheric 

deposits, system design, and operating experience. Industry guidance 

on evaluating susceptibility has been published by EPRI (Fuhr et al., 2015).

Justification for not conducting inspections for localized corrosion or stress 

corrosion cracking should be provided on a case-by-case basis for each 

ISFSI site where welded stainless steel canisters are in use. Acceptable 

justification may be based on a comparison of susceptibility for the ISFSI 

location versus at least two other ISFSI sites determined to have greater 

susceptibility but showed no evidence of localized corrosion or stress 

corrosion cracking in inspections completed within 5 years of the time of the 

assessment. The justification must consider the full range of available ISFSI 

susceptibility assessments and welded stainless steel canister examination 

results.

Data Collection

Canister Examination: Documentation of the examination of the canister, 

location, and appearance of deposits. Assessment of the suspect areas 

where corrosion products were observed as described in corrective actions.

Bounding Analysis: A complete listing of other sites considered, 

susceptibility assessments for those sites and results of examinations 

conducted at those sites. Justification for not including other sites where 

examinations showed evidence of localized corrosion and/or stress 

corrosion cracking.

Frequency

Once every 5 years
B-6

Element Description

4. Detection of Aging 

Effects (cont’d)

Timing of Inspections

The timing of the inspections includes the pre-application inspection or 

general-licensee baseline inspection, performed per Sections 3.4.1.2 and 

3.6.1.10 of this NUREG, and at the frequency specified by the AMP.

Alternative detection methods or techniques may be provided. For 

these cases:

• The method or technique should be adequate and proven to be capable 

of evaluating the condition of the external surface of the canister against 

the acceptance criteria for the detection of localized corrosion and stress 

corrosion cracking.

• The proposed intervals for inspection or monitoring are consistent 

with applicable site-specific, design-specific, or industrywide operating 

experience and should have sufficient frequency to ensure that the 

confinement function will be maintained until the next scheduled 

inspection.

• The data collection methods should be sufficient for evaluating localized 

corrosion and stress corrosion cracking and should reference specific 

methods to be used for data acquisition including any applicable 

consensus codes and standards.

5. Monitoring and 

Trending

Monitoring and trending methods are in accordance with ASME Section XI 

evaluation criteria.

Monitoring and trending methods reference plans/procedures used to:

• Establish a baseline before or at the beginning of the period of extended 

operation.

• Track trending of parameters or effects not corrected following a 

previous inspection, including:

– the locations and size of any areas of localized corrosion or the 

stress corrosion cracking

– the disposition of canisters with identified aging effects and the 

results of supplemental canister inspections

Monitoring and trending should also include:

• documentation of the appearance of the canister, particularly at 

welds and in crevice locations, with images and video that will allow 

comparison in subsequent examinations

• changes to the size and number of any rust colored stains as a result of 

iron contamination of the surface in subsequent inspections
B-7

6. Acceptance Criteria No indications of localized corrosion pits, etching, crevice corrosion, stress 

corrosion cracking, red-orange colored corrosion products emanating from 

crevice locations, or red-orange colored corrosion products in the vicinity of 

canister fabrication welds, closure welds, and welds associated with 

temporary attachments during canister fabrication.

Confirmed or suspected areas of crevice corrosion, pitting corrosion and 

stress corrosion cracking must be assessed in accordance with acceptance 

standards identified in ASME Section XI, IWB-3514. Flaws exceeding the 

acceptance standards in IWB-3514.1 must be evaluated using the 

acceptance criteria identified in IWB-3640.

Indications Requiring Additional Evaluation

Although shop and handling procedures include controls to prevent iron 

contamination of the stainless steel surfaces, contamination does occur and 

is usually identified by rust-colored surface deposits. Iron contamination can 

exacerbate CISCC in stainless steels. In accessible locations, removal of 

the deposits and rust stains that reveal undamaged welds (i.e., absence of 

pits, crack, localized attack, or etching) and the original machining/grinding 

marks on the stainless steel base metal, including weld heat affected zones, 

may be used to confirm that localized corrosion or stress corrosion cracking 

have not been initiated.

Indications of interest that are subject to additional examination and 

disposition include:

• localized corrosion pits, crevice corrosion, stress corrosion cracking, 

and etching (note that these indications may be covered by obstructions 

(i.e., crevices)); deposits; or corrosion products

• discrete red-orange colored corrosion products that are 1 mm in 

diameter or larger especially those adjacent to fabrication welds, closure 

welds, locations where temporary attachments may have been welded 

to and subsequently removed from the stainless steel dry storage 

canister, and the weld heat affected zones of these areas

• linear appearance of any color of corrosion products of any size parallel 

to or traversing fabrication welds, closure welds, locations where 

temporary attachments may have been welded to and subsequently 

removed from the stainless steel dry storage canister, and the weld heat 

affected zones of these areas

• red-orange colored corrosion products greater than 1 mm in diameter 

combined with deposit accumulations in any location of the stainless 

steel canister

• red-orange colored corrosion tubercles of any size

• red-orange corrosion products present at the mouth of a crevice that 

includes a portion of the canister surface

Alternative acceptance criteria may be provided. For such cases, the 

acceptance criteria should:

• Include a quantitative basis (justifiable by operating experience, 

engineering analysis, consensus codes/standards).

• Avoid use of non-quantifiable phrases (e.g., significant, moderate, minor, 

little, slight, few).

• Be achievable and clearly actionable.
B-8

Element Description

7. Corrective Actions The corrective actions are in accordance with the specific or general 

licensee Quality Assurance (QA) Program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B. The QA Program ensures that 

corrective actions are completed within the specific or general licensee’s 

Corrective Action Program (CAP), and include provisions to:

• Perform functionality assessments.

• Perform apparent cause evaluations, and root cause evaluations.

• Address the extent of condition.

• Determine actions to prevent recurrence for significant conditions 

adverse to quality; ensure justifications for non-repairs.

• Trend conditions.

• Identify operating experience actions, including modification to the 

existing AMP (e.g., increased frequency).

• Determine if the condition is reportable per 10 CFR 72.75.

Example provisions are provided below, for assessment of extent of 

condition and evaluation criteria for canisters with aging effects, following 

from the example acceptance criteria. Alternative corrective actions to those 

provided below may be appropriate, per the licensees CAP evaluation to 

determine appropriate corrective actions to be taken if AMP acceptance 

criteria are not met. 

Extent of Condition

Confirmation of localized corrosion or stress corrosion cracking may warrant 

inspection of additional canisters at the same ISFSI location to determine 

the extent of condition. Priority for additional inspections should be to 

canisters with similar time in service and initial loading. Canisters with 

confirmed localized corrosion or stress corrosion cracking must be evaluated 

for continued service. Canisters with localized corrosion or stress corrosion 

cracking that do not meet the prescribed evaluation criteria must be repaired 

or replaced.
B-9

Element Description

7. Corrective Actions 

(cont’d)

Evaluation Criteria for Canisters with Aging Effects

For austenitic stainless steel canisters covered by an AMP that utilizes the 

inspection and acceptance criteria in ASME B&PV code Section XI for 

Class 1 piping system, the disposition of canisters should be commensurate 

with in-service inspection results:

• Canisters with no evidence of corrosion are permitted to remain 

in service and will continue to be evaluated at 5-year intervals.

• Canisters with rust deposits that are determined to be a result of iron 

contamination but do not have evidence of localized corrosion or stress 

corrosion cracking are permitted to remain in service and will continue to 

be evaluated at 5-year intervals.

• Canisters that show evidence of localized corrosion or stress corrosion 

cracking that does not exceed the acceptance standards in IWB-3514.1 

are permitted to remain in service and will be evaluated at 5-year 

intervals. Sample size should be increased to assess canisters with 

similar susceptibility assessments.

• Canisters that show evidence of localized corrosion or stress corrosion 

cracking that exceeds the acceptance standards in IWB-3514.1 but 

meet the acceptance criteria identified in IWB-3640 including the 

required evaluation per IWB-3641(a) using the prescribed evaluation 

procedures, are permitted to remain in service and should be evaluated 

at 3-year intervals. Sample size should be increased to assess 

canisters with similar susceptibility assessments.

• Canisters that show evidence of localized corrosion or stress corrosion 

cracking that exceeds acceptance criteria identified in IWB-3640 are not 

permitted to remain in service without an engineering analysis and/or 

mitigation actions. Sample size should be increased to assess canisters 

with similar susceptibility assessments.

8. Confirmation 

Process

The confirmation process will be commensurate with the specific or general 

licensee Quality Assurance (QA) Program approved under 10 CFR Part 72, 

Subpart G or 10 CFR Part 50, Appendix B. The QA Program ensures that 

the confirmation process includes provisions to preclude repetition of 

significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• Determine follow-up actions to verify effective implementation 

of corrective actions.

• Monitor for adverse trends due to recurring or repetitive findings 

or observations.
B-10

Element Description

9. Administrative 

Controls

The administrative controls are in accordance with the specific- or 

general-licensee QA Program approved under 10 CFR Part 72, Subpart G, 

or 10 CFR Part 50, Appendix B, respectively. The QA Program ensures that 

administrative controls include provisions that define:

• instrument calibration and maintenance

• inspector requirements

• record retention requirements

• document control

The administrative controls describe or reference:

• methods for reporting results to the NRC per 10 CFR 72.75

• frequency for updating AMP based on site-specific, design-specific, 

and industrywide operating experience
B-11

Element Description

10. Operating 

Experience

The AMP references and evaluates applicable operating experience, both

before renewal and will continue to do so as new operating experience is 

developed and made available after renewal, including:

• internal and industrywide condition reports

• internal and industrywide corrective action reports

• vendor-issued safety bulletins

• NRC Generic Communications

• applicable DOE or industry initiatives (e.g., EPRI- or DOE-sponsored 

inspections)

The AMP clearly identifies any degradation in the referenced operating 

experience as either age-related or event-driven, with proper justification for 

that assessment. Past operating experience supports the adequacy of the 

proposed AMP, including the method/technique, acceptance criteria, and 

frequency of inspection.

The AMP references the methods for capturing operating experience from 

other ISFSIs with similar in-scope SSCs.

CISCC of austenitic stainless steels is a known degradation mechanism 

for aqueous environments; however, operating experience in aqueous 

environments is not directly applicable in assessing the potential for 

atmospheric CISCC for austenitic stainless steel dry storage canisters. 

Atmospheric CISCC of austenitic stainless steels has been reported in a 

range of industries including welded stainless steel components and piping 

in operating nuclear power plants.

Spent Fuel Storage

Inspections of dry storage canisters after 20 years in service have been 

conducted at a few independent spent fuel installation (ISFSI) sites. Details 

of the inspection conducted at the Calvert Cliffs nuclear power plant ISFSI 

are documented in a recent EPRI report (Waldrop et al., 2014; Bryan and 

Enos, 2015). No evidence of localized corrosion was identified but some 

amount of chloride-containing salts were determined to be present and 

corrosion products believed to be related to iron contamination were 

identified.

Operating Power Reactors

NRC Information Notice 2012-20 (NRC, 2012) documents previous cases 

of atmospheric CISCC of welded stainless steel piping systems and tanks at 

operating reactor locations. Atmospheric CISCC growth rates determined 

from operating experience at both domestic and foreign nuclear power 

plants including events at San Onofre, Turkey Point, St. Lucie, and Koeberg 

(South Africa) range from 3.6 × 10-12 m/s to 2.9 × 10-11 m/s for components 

at ambient temperatures.
B-12

Element Description

10. Operating 

Experience (cont’d)

Relevant Literature and Testing

Electric Power Research Institute (EPRI) has recently conducted a literature 

review of CISCC which summarizes the results of many previous laboratory 

investigations (Gorman et al., 2014).

The NRC has recently published the results of a completed investigation of 

CISCC testing of type 304, 304L and 316L stainless steel and welds (He, 

et al., 2014). This study indicates that SCC was initiated at stresses just 

above the yield strength in tests conducted using 304 stainless steel C-ring 

specimens. Testing with U-bend specimens showed that CISCC was 

observed with the lowest simulated sea salt concentrations tested (100 mg 

salt/m2 or ~55 mg chloride/m2) at temperatures of 52°C (125.6°F) using a 

maximum absolute humidity of 30 g/m3, which is generally accepted as 

being near the maximum absolute humidity in a natural environment.

Both laboratory and field investigations have been conducted by Central 

Research Institute of Electric Power Industry (CRIEPI) and Tokyo Electric 

Power Company (TEPCO). This includes the early work by Tokiwai et al. 

(1985) who reported the critical surface chloride concentrations of 8 mg/m2 

for CISCC on sensitized Type 304 stainless steel. Kosaka (2008) reported 

crack growth rates of 9.6 × 10-12 m/s obtained in natural exposure tests on 

Miyakojima Island with Type 304 base metals and welds, Type 304L welds 

and Type 316LN welds. Hayashibara, et al. (2008) reported activation 

energy for crack growth in Type 304 stainless steel of 5.6 to 9.4 kcal/mol (23 

to 39 kJ/mol) based on testing conducted at temperatures of 50 to 80°C (122 

to 176°F).
B-13

Element Description

References

American Society of Mechanical Engineers, “Boiler and Pressure Vessel 

Code Section XI—Rules for Inservice Inspection of Nuclear Power Plant 

Components,” New York, NY: ASME, 2007.

Bryan, C.R., D.G. Enos, “Analysis of Dust Samples Collected from Spent 

Nuclear Fuel Interim Storage Containers at Hope Creek, Delaware, and 

Diablo Canyon, California,” SAND2014-16383, Albuquerque, NM: Sandia 

National Laboratories, July 2014.

Cumblidge, S.E., M.T. Anderson, S.R. Doctor, “An Assessment of Visual 

Testing,” NUREG/CR 6860, Pacific Northwest National Laboratory, 

Richland, WA, 2004, ADAMS Accession No. ML043630040.

Cumblidge, S.E., et al., “A Study of Remote Visual Methods to Detect 

Cracking in Reactor Components,” NUREG/CR-6943, Pacific Northwest 

National Laboratory, Richland, WA, 2007, ADAMS Accession 

No. ML073110060.

Fuhr, K., J. Broussard, G. White, “Susceptibility Assessment Criteria for 

Chloride-Induced Stress Corrosion Cracking (CISCC) of Welded Stainless 

Steel Canisters for Dry Cask Storage Systems,” Electric Power Research 

Institute (EPRI), EPRI-3002005371, Palo Alto, CA, 2015.

Gorman, J., K. Fuhr, J. Broussard, “Literature Review of Environmental 

Conditions and Chloride-Induced Degradation Relevant to Stainless Steel 

Canisters in Dry Cask Storage Systems,” EPRI-3002002528, Palo Alto, 

CA, 2014.

Hayashibara, H., et al. “Effect of Temperature and Humidity on Atmospheric 

Stress Corrosion Cracking of 304 Stainless Steel,” NACE International, 

Corrosion 2008, paper 08492, Houston, TX, 2008.

He, X., et al., “Assessment of Stress Corrosion Cracking Susceptibility 

for Austenitic Stainless Steels Exposed to Atmospheric Chloride and 

Non-Chloride Salts,” NUREG/CR-7170, U.S. Nuclear Regulatory 

Commission, February 2014, ADAMS Accession No. ML14051A417.

Kosaki, A., “Evaluation Method of Corrosion Lifetime of Conventional 

Stainless Steel Canister under Oceanic Air Environment,” Nuclear 

Engineering and Design, Vol. 238, 2008, pp.1233–1240.

Selby, G., “BWR Vessel and Internals Project, Reactor Pressure Vessel 

and Internals Examination Guidelines,” EPRI 1011689, TR-105696-R8 

(BWRVIP-03) Rev. 8, Palo Alto, CA, 2005.

Tokiwai, M., H. Kimura, H. Kusanagi, “The Amount of Chlorine 

Contamination for Prevention of Stress Corrosion Cracking in Sensitized 

Type 304 Stainless Steel,” Corrosion Science, Vol. 25 Issue 8–9, 1985, 

pp. 837–844.
B-14

Element Description

References (cont’d) U.S. Nuclear Regulatory Commission (NRC), Information Notice 2012-20: 

“Potential Chloride-Induced Stress Corrosion Cracking of Austenitic 

Stainless Steel and Maintenance of Dry Cask Storage System Canisters,” 

November 14, 2012, ADAMS Accession No. ML12319A440.

Waldrop, K. et al., “Calvert Cliffs Stainless Steel Dry Storage 

Canister Inspection,” EPRI-1025209, EPRI, Palo Alto, CA, 2014.
B-15

Example AMP for Reinforced Concrete Structures

An example aging management program (AMP) for reinforced concrete structures is provided 

below. The AMP consists of condition monitoring, performance monitoring, mitigation and 

prevention activities. The program includes periodic visual inspections by personnel qualified 

to monitor reinforced concrete for applicable aging effects, such as those described in the 

American Concrete Institute (ACI) guides 349.3R-02, ACI 201.1R-08, and American National 

Standards Institute/American Society of Civil Engineers guidelines (ANSI/ASCE) 11-99. 

Identified aging effects are evaluated against acceptance criteria derived from the design bases 

or industry guides and standards, including ACI 349, ACI 318, ACI 349.3R-02 and the American 

Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 

Subsection IWL.

The program also includes periodic sampling and testing of groundwater and the need to 

assess the impact of any changes in its chemistry on below-grade concrete structures. 

Additional activities include radiation surveys to ensure the shielding functions of the concrete 

structure are maintained and daily inspections to ensure the air convection vents are not 

blocked, if applicable (per the requirements of the approved design bases). The program also 

includes provisions where modifications may be appropriate for specific license renewals.
B-16

Table B-2. Example AMP for Reinforced Concrete Structures

Element Description

1. Scope of Program The scope of the program includes the following aging 

management activities:

1. visual inspection of above-grade (accessible, inaccessible) and 

below-grade (underground) concrete areas (See Element 4 for sample 

size and justification of areas to be inspected)

2. groundwater chemistry monitoring program to identify conditions 

conducive to below-grade (underground) aging mechanisms:

• corrosion of embedded steel

• chemical attack (chloride and sulfate induced degradation

3. radiation surveys to:

• Ensure compliance with 10 CFR 72.104 (i.e., dose equivalent 

requirements beyond the controlled area during normal and 

off-normal conditions of storage).

• Monitor performance of the concrete as a neutron/gamma shield 

at near system locations as an indicator of concrete degradation.

The program provides means to address the following aging effects and 

mechanisms, as described in ACI 349.3R-02 and ASCE/SEI 11-99:

• cracking or loss of material (spalling, scaling) due to freeze-thaw 

degradation

• cracking or loss of material (spalling, scaling) due to chemical attack 

(chloride, sulfate induced)

• cracking and loss of strength due to cement aggregate reactions

• cracking, loss of material, and loss of bond due to corrosion of 

embedded steel

• increase in porosity/permeability and loss of strength due to leaching of 

calcium hydroxide (Ca(OH)2)

• cracking and distortion due to long-term settlement

• cracking and reduction in strength due to gamma and neutron irradiation

Calculations or analyses (time-limited aging analyses, when appropriate) 

may be used to demonstrate that aging effects due to irradiation do not 

require an AMP. More specifically, the renewal application may demonstrate 

that no part of the concrete exceeds critical cumulative fluences of 

1017 neutrons/m2 or 1010 rad (gamma dose), per ACI 349.3R-02.a

Additional site-specific AMPs may be required for the following scenarios:

• A dewatering system is used to prevent long-term settlement.

• The design bases includes embedded aluminum subcomponents 

without a protective insulating coating.

• Protective coatings are relied upon to manage the effects of aging for 

a subcomponent.
B-17

Element Description

2. Preventive Actions Preventive actions include (1) continuance of inspections to ensure that 

air inlet and outlet vents are not blocked, or (2) temperature monitoring, if 

applicable, to ensure design temperature limits are not exceeded. These 

inspections would be part of the approved design bases and be continued 

for the sample size and inspection frequency identified in the respective 

technical specification.

Additional preventive actions are not required for structures designed and 

fabricated in accordance to ACI 318 or ACI 349, as specified in the design 

bases. Otherwise, a site-specific AMP may be required.

3. Parameters 

Monitored or 

Inspected

For visual inspections, the parameters monitored/inspected quantify the 

following aging effects:

• cracking

• material loss (spalling, scaling)

• loss of bond

• increased porosity/permeability

The AMP references the following parameters for characterizing the above 

aging effects, as appropriate:b

• affected surface area

• geometry/depth of defect

• cracking, crazing, delaminations, drummy areas

• curling, settlements or deflections

• honeycombing, bug holes

• popouts and voids

• exposure of embedded steel

• staining/ evidence of corrosion

• dusting, efflorescence of any color

The parameters evaluated consider any surface geometries that may 

support water ponding and potentially increase the rate of degradation.

For the groundwater chemistry program, the parameters 

monitored/inspected include:

• water pH

• concentration of chlorides and sulfates in the water

For radiation surveys, the parameters monitored/inspected include gamma 

dose rate and/or neutron fluence rate.
B-18

Element Description

4. Detection of 

Aging Effects

Method/Technique

The method/technique achieves the acceptance criteria, as defined in 

AMP Element 6. An engineering justification or technical bases is provided, 

which references applicable consensus guides, codes and standards, or 

calibration procedures that ensure the method or technique will provide 

reliable data.

For visual inspections, the method/technique is defined as:

• visual method for normally accessible areas (e.g., feeler gauges, 

crack comparators)

• visual method for normally inaccessible areas (site-qualified remote 

inspection system)

Procedures for remote visual inspections should be demonstrated to ensure 

the acceptance criteria in ACI 349.3R is achievable; procedure attributes 

should include equipment resolution, lighting requirements, etc. and 

reference applicable standards when possible.

For the groundwater chemistry program, the method/technique is defined as 

a chemical analysis method with a valid measurement range relative to the 

acceptance criteria.

For radiation surveys, the method/technique is defined as calibrated neutron 

and gamma detectors with valid energy ranges.

Frequency of Inspection

The proposed inspection schedule is commensurate with ACI 349.3R-02. 

Alternative inspection frequencies provide a valid technical basis 

(engineering justification, operating experience data) for any deviation from 

ACI 349.3R-02.

For visual inspections, the frequency of inspection is defined as:

• for above-grade (accessible and inaccessible) areas: ≤ 5 years

• for below-grade (underground) areas: ≤ 10 years, and when excavated 

for any reason

• the use of opportunistic inspections in lieu of planned inspections per the 

above schedule provides a valid technical basis (engineering 

justification, operating experience data).

For the groundwater chemistry program, the frequency of monitoring is 

justified (e.g., quarterly, semiannual).

For radiation surveys, the frequency of monitoring is justified 

(e.g., quarterly).
B-19

Element Description

4. Detection of Aging 

Effects (cont’d) Sample Size

Visual inspections cover 100 percent of readily accessible surfaces of all 

reinforced concrete structures in operation (e.g., all normally accessible 

exterior surfaces of all loaded overpacks), and 100 percent of normally 

inaccessible surfaces for a justified subset of the reinforced concrete 

structures in operation (e.g., interior surfaces of two overpacks, including the 

overpack earliest loaded and the overpack loaded with the highest heat-load 

canister). The extent of inspection coverage should be specified 

and demonstrated to sufficiently characterize the condition of the structure.

For the groundwater chemistry program and radiation surveys, the sample 

size clearly identifies and justifies specific locations where 

inspection/monitoring will be conducted to sufficiently characterize the 

condition of the structure (e.g., periodic dose rate measurements will be 

performed at same locations specified in the technical specifications for dose 

rate measurements at loading).

Data Collection:

Data collection for visual inspections is commensurate with applicable 

consensus codes/standards/guides:

• for example, ACI 224.1R for quantitative analysis (crack width, extent), 

ACI 562, ACI 364.1R.

The AMP references a clearinghouse for documenting inspection/monitoring 

operating experience.

Timing

The timing of the inspections includes the pre-application inspection or 

general-licensee baseline inspection, performed per Sections 3.4.1.2 and 

3.6.1.10 of this NUREG, and at the frequency justified by the AMP.

5. Monitoring 

and Trending

Monitoring and trending methods are commensurate with defect evaluation 

guides and standards (e.g., ACI 201.1R, ACI 207.3R, ACI 364.1R, ACI 562, 

or ACI 224.1R for crack evaluation).

Monitoring and trending methods reference plans/procedures used to:

• Establish a baseline before or at the beginning of the period of 

extended operation.

• Track trending of parameters or effects not corrected in a previous 

inspection, for example:

• crack growth/extent

• pore/void density and affected areas

• dose rates
B-20

Element Description

6. Acceptance Criteria For visual inspections, the acceptance criteria are commensurate with the 

three-tier quantitative criteria in ACI 349.3R-02:

• acceptance without further evaluation

• acceptance after review

• acceptance requiring further evaluation

The acceptance criteria clearly identify when a condition is to be entered 

in the Corrective Action Program (e.g., when Tier 2 acceptance per 

ACI 349.3R-02 is exceeded).

For the groundwater chemistry program, the acceptance criteria are

commensurate with ASME Code Section XI, Subsection IWL, which 

states that an aggressive below-grade environment is defined as:

• pH < 5.5, chlorides > 500 ppm, or sulfates > 1500 ppm

For radiation surveys, the acceptance criteria are justified and sufficient 

to ensure compliance with 10 CFR 72.104 and identify dose rates that 

statistically exceed calculated/expected dose rates at pre-determined 

measurement locations. The adequacy of the acceptance criteria considers 

measured dose rates versus calculated/expected dose rates for a dry 

storage system (DSS) given the DSS contents and accounting for the decay 

of the source term since the DSS loading. Measurement locations should be 

consistent with those specified in the license/CoC conditions/technical 

specification (if any) and/or locations where dose rates were calculated 

in the FSAR and likely measured at the time of loading.

Alternative acceptance criteria may be provided. For such cases, the 

acceptance criteria should:

• Include a quantitative basis (justifiable by operating experience, 

engineering analysis, consensus codes/standards).

• Avoid use of non-quantifiable phrases (e.g., significant, moderate, minor, 

little, slight, few).

• Be achievable and clearly actionable.

7. Corrective Actions The corrective actions are in accordance with the specific- or 

general-licensee QA Program approved under 10 CFR Part 72, Subpart G, 

or 10 CFR Part 50, Appendix B, respectively. The QA Program ensures that 

corrective actions are completed within the specific or general licensee’s 

CAP, and include provisions to:

• Perform functionality assessments.

• Perform apparent cause evaluations, and root cause evaluations.

• Address the extent of condition.

• Determine actions to prevent recurrence for significant conditions 

adverse to quality; ensure justifications for non-repairs.

• Trend conditions.

• Identify operating experience actions, including modification to the 

existing AMP (e.g., increased frequency).

• Determine if the condition is reportable to the NRC per 10 CFR 72.75.

The AMP references applicable concrete rehabilitation guides or standards, 

for example:

• cracking: ACI 224.1R, ACI 562, ACI 364.1R, and ACI RAP Bulletins

• spalling/scaling: ACI 562, ACI 364.1R, ACI 506R, and ACI RAP 

Bulletins
B-21

Element Description

8. Confirmation 

Process

The confirmation process is commensurate with the specific- or 

general-licensee QA Program approved under 10 CFR Part 72, Subpart G, 

or 10 CFR Part 50, Appendix B, respectively. The QA Program ensures that 

the confirmation process includes provisions to preclude repetition of 

significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• determine follow-up actions to verify effective implementation 

of corrective actions

• monitor for adverse trends due to recurring or repetitive findings

or observations

9. Administrative 

Controls

The administrative controls are in accordance with the specific- or 

general-licensee QA Program approved under 10 CFR Part 72, Subpart G, 

or 10 CFR Part 50, Appendix B, respectively. The QA Program ensures that 

the administrative controls include provisions that define:

• instrument calibration and maintenance

• inspector requirements (commensurate with ACI 349.3R-02)

• record retention requirements

• document control

The administrative controls describe or reference:

• methods for reporting results to the NRC per 10 CFR 72.75

• frequency for updating the AMP based on industrywide 

operating experience

10. Operating 

Experience

The AMP references and evaluates applicable operating experience, both 

before renewal and will continue to do so as new operating experience is 

developed and made available after renewal, including:

• internal and industrywide condition reports

• internal and industrywide corrective action reports

• vendor-issued safety bulletins

• NRC Generic Communications

• applicable DOE or industry initiatives (e.g., EPRI- or DOE-sponsored 

inspections)

The AMP clearly identifies any degradation in the referenced operating 

experience as either age-related or event-driven, with proper justification for 

that assessment. Past operating experience supports the adequacy of the 

proposed AMP, including the method/technique, acceptance criteria, and 

frequency of inspection.

The AMP references the methods for capturing operating experience from 

other ISFSIs with similar in-scope SSCs.
B-22

Element Description

References ACI 201.1R-08, “Guide for Conducting a Visual Inspection of Concrete in 

Service,” 2008.

ACI 207.3R-94, “Practices for Evaluation of Concrete in Existing Massive 

Structures for Service Conditions,” 2008.

ACI 224.1R-07, “Causes, Evaluation, and Repair of Cracks in Concrete 

Structures,” 2007.

ACI 318-05, “Building Code Requirements for Structural Concrete and 

Commentary,” 2005.

ACI 349.3R-02, “Evaluation of Existing Nuclear Safety-Related Concrete 

Structures,” 2010.

ACI 349-06, “Code Requirements for Nuclear Safety-Related Concrete 

Structures and Commentary,” 2007.

ACI 364.1R-07, “Guide for Evaluation of Concrete Structures before 

Rehabilitation,” 2007.

ACI 506R-05, “Guide to Shotcrete,” 2005.

ACI 562-13, “Code Requirements for Evaluation, Repair, and Rehabilitation 

of Concrete Buildings,” 2013.

ACI CT-13, “ACI Concrete Terminology,” 2013.

ASCE/SEI 11-99, “Guideline for Structural Condition Assessment of Existing 

Buildings,” 2000.

ASME Boiler and Pressure Vessel Code, Section XI, Subsection IWL, 

“Requirements for Class CC Concrete Components of Light-Water-Cooled 

Plants,” 2013.

NRC, “Standard Review Plan for Spent Fuel Dry Storage Facilities,” 

NUREG-1567, Rev. 0. Washington, DC, 2000. ADAMS Accession 

No. ML003686776.

NRC, “Standard Review Plan for Spent Fuel Dry Storage Systems at a 

General License Facility,” NUREG-1536, Rev. 1, Washington, DC, 2010a. 

ADAMS Accession No. ML101040620.

NRC, “Generic Aging Lessons Learned (GALL) Report,” NUREG-1801,

Rev. 2, Washington, DC, 2010c. ADAMS Accession No. ML103490041.

a The staff recognizes the critical cumulative fluence in ACI 349.3R-02 as adequately conservative. The 

applicant may choose to justify a higher critical cumulative fluence if a technical bases is provided. b Terminology consistent with ACI standard CT-13.
B-23

Example of a High Burnup Fuel Monitoring and Assessment Program

An example of a High Burnup (HBU) Fuel 1 Monitoring and Assessment Program is provided 

below. This is a licensee program that monitors and assesses data and other information 

regarding HBU fuel performance, to confirm that the design-bases HBU fuel configuration is 

maintained during the period of extended operation. This example HBU Monitoring and 

Assessment Program relies on a surrogate demonstration program to provide data on HBU fuel 

performance. Guidance for determining if a surrogate demonstration program can provide the 

data to support a licensee’s HBU Fuel Monitoring and Assessment Program is given in 

Appendix D. Although this example focuses on the use of a surrogate demonstration program, 

a licensee may use alternative approaches that are appropriately justified.

The aging management review is not expected to identify any aging effects that could lead to 

fuel reconfiguration, as long as the HBU fuel is stored in a dry inert environment, temperature 

limits are maintained, and thermal cycling is limited (see Sections 2.4.2.1 and 3.4.1.4). 

Short-term testing (i.e., laboratory scale testing up to a few months) and scientific analyses 

examining the performance of HBU fuel have provided a foundation for the technical basis that 

storage of HBU fuel in the period of extended operation may be performed safely and in 

compliance with regulations. However, there has been relatively little operating experience, 

to date, with dry storage of HBU fuel.

Therefore, the purpose of a HBU Fuel Monitoring and Assessment Program is to monitor and 

assess data and other information regarding HBU fuel performance to confirm there is no 

degradation of HBU fuel that would result in an unanalyzed configuration during the period 

of extended operation. The following description of an example HBU Fuel Monitoring and 

Assessment Program presents the applicable information in a format using each element of 

an effective aging management program, to provide a framework for such a monitoring and 

assessment program.

 1 Fuel assemblies with discharge burnup greater than 45 gigawatt-days per metric ton of uranium (GWd/MTU)
B-24

Table B-3. Example of a High Burnup Fuel Monitoring and Assessment Program

Element Description

1. Scope of 

the Program

The scope of the program provides a description of: (1) the design-bases 

characteristics of the HBU fuel, (2) the surrogate demonstration program that 

will be used to provide data on the applicable design-bases HBU fuel 

performance, and (3) how the parameters of the surrogate demonstration 

program are applicable to the design-bases HBU fuel.

Aging effects will be determined for material/environment combinations 

through an alternative surrogate demonstration program meeting the 

guidance in Appendix D.

Example language to address this “scope of the program” element follows:

Fuel stored in a [define cask/canister model] is limited to an assembly 

average burnup of [define design-bases limit] GWd/MTU. The cladding 

materials for the HBU fuel are [define types of cladding], and the fuel is 

stored in a dry helium environment. HBU fuel was first placed into dry 

storage in a [define cask/canister model] on [start date of storage term 

of first storage of HBU fuel].

The program relies on the joint Electric Power Research Institute (EPRI) 

and Department of Energy (DOE) “HBU Dry Storage Cask Research and 

Development Project” (HDRP) (EPRI 2014), conducted in accordance with 

the guidance in Appendix D, as a surrogate demonstration program that 

monitors the performance of HBU fuel in dry storage.

The HDRP is a program designed to collect data from a spent nuclear fuel 

storage system containing HBU fuel in a dry helium environment. The 

program entails loading and storing a AREVA TN-32 bolted lid cask (the 

“Research Project Cask”) at Dominion Virginia Power’s North Anna Power 

Station with intact HBU fuel (of average assembly burnups ranging between 

53 GWd/MTU and 55.5 GWd/MTU). The fuel to be used in the program 

include four kinds of cladding (Zircaloy-4, low-tin Zircaloy-4, ZirloTM, and 

M5TM). The Research Project Cask is to be licensed to the temperature 

limits contained in Interim Staff Guidance (ISG) 11, Rev. 3 (NRC 2003), and 

loaded such that the fuel cladding temperature is as close to the limit as 

practicable. [If an alternative surrogate demonstration program is used, 

provide a description of the program.]
B-25

Element Description

1. Scope of the 

Program (cont’d)

The parameters of the surrogate demonstration program are applicable to 

the design-bases HBU fuel, as the: (1) maximum assembly-average burnup 

of the design-bases HBU fuel [define value] is less than the 

assembly-average burnup of the fuel in the surrogate demonstration 

program [define value]; (2) the cladding type of the design-bases HBU fuel 

[define type] is the same as the surrogate demonstration program [define 

type]; and (3) the peak temperatures in the surrogate demonstration 

program [define values] bound the peak temperatures of the loaded systems 

(or design bases) [define values].

a

2. Preventive Actions There are no specific preventive actions associated with this HBU Fuel 

Monitoring and Assessment Program. However, the applicant should 

discuss the design-bases characteristics of the licensed/certified dry storage 

system, in terms of initial cask loading operations, to show the HBU fuel is 

stored in a dry inert environment.

Example language follows:

During the initial loading operations of the cask/canister, the design and 

ISFSI Technical specifications (TS) require that the fuel be stored in a dry 

inert environment. TS [name and number] demonstrates that the 

cask/canister cavity is dry by maintaining a cavity absolute pressure less 

than or equal to [value] for a [time period] with the cask/canister isolated 

from the vacuum pump. TS [name and number], requires that the 

cask/canister then be backfilled with helium. These two TS requirements 

ensure that the HBU fuel is stored in an inert environment thus preventing 

cladding degradation due to oxidation mechanisms. TS [name and number]

also requires that the helium environment be established within [time] hours 

of commencing cask/canister draining. The cask/canister is loaded in 

accordance with the criteria of ISG-11.

3. Parameters 

Monitored or 

Inspected

The applicant identifies the parameters monitored and inspected in a 

surrogate demonstration program that are applicable to its particular 

design-bases HBU fuel and describes how this meets the guidance of 

Appendix D.

4. Detection of 

Aging Effects

The applicant identifies the detection of aging effects in a surrogate 

demonstration program that are applicable to its particular design-bases 

HBU fuel and describes how this meets the guidance of Appendix D.
B-26

Element Description

5. Monitoring 

and Trending

As information/data from a surrogate demonstration program or from 

other sources (such as testing or research results and scientific analyses) 

becomes available, the licensee will monitor, evaluate, and trend the 

information via its operating experience program and/or the Corrective 

Action Program (CAP) to determine what actions should be taken.

The licensee will evaluate the information/data from a surrogate 

demonstration program or from other sources to determine whether 

the acceptance criteria in Element 6 are met.

• If all of the acceptance criteria are met, no further assessment 

is needed.

• If any of the acceptance criteria are not met, the licensee must conduct 

additional assessments and implement appropriate corrective actions 

(see Element 7).

Formal evaluations of the aggregate information from a surrogate 

demonstration program and other available domestic or international 

operating experience (including data from monitoring and inspection 

programs, NRC generic communications, and other information) will be 

performed at specific points in time during the period of extended operation, 

as delineated in Table B-4.

6. Acceptance Criteria The High Burnup Fuel Monitoring and Assessment Program acceptance 

criteria are:

• Hydrogen content—maximum hydrogen content of the cover gas over 

the approved storage period should be extrapolated from the gas 

measurements to be less than the design-bases limit for 

hydrogen content.

• Moisture content—the moisture content in the cask/canister, accounting 

for measurement uncertainty, should be less than the expected upper 

bound moisture content per the design-bases drying process.

b

• Fuel condition/performance—nondestructive examination (e.g., fission 

gas analysis) and destructive examination (e.g., to obtain data on creep, 

fission gas release, hydride reorientation, cladding oxidation, and 

cladding mechanical properties) should confirm the design-bases fuel 

condition (i.e., no changes to the analyzed fuel configuration considered 

in the safety analyses of the approved design bases).

The applicant should provide information on the design-bases 

characteristics of the dry storage system or ISFSI, with regard to these 

criteria. The applicant should reference the source of specific values, or 

explain any assumptions made, for defining design-bases characteristics 

of the fuel condition/performance.
B-27

Element Description

7. Corrective Actions The corrective actions are in accordance with the specific- or 

general-licensee QA Program approved under 10 CFR Part 72, Subpart G, 

or 10 CFR Part 50, Appendix B, respectively.

Corrective actions should be implemented if data from a surrogate 

demonstration program or other sources of information indicate that any 

of the High Burnup Fuel Monitoring and Assessment Program acceptance 

criteria (in Element 6) are not met.

If any of the acceptance criteria are not met, the licensee will:

(1) Assess fuel performance (effects on fuel and changes to 

fuel configuration).

(2) Assess the design-bases safety analyses, considering degraded fuel 

performance (and any changes to fuel configuration), to determine the 

ability of the dry storage system/ISFSI to continue to perform its 

intended functions under normal, off-normal, and accident conditions.

The licensee will determine what corrective actions should be taken to:

(1) Manage fuel performance, if any.

(2) Manage impacts related to degraded fuel performance to ensure that all 

intended functions for the dry storage system/ISFSI are met.

In addition, the licensee will obtain the necessary NRC approval in the 

appropriate licensing/certification process for modification of the design 

bases to address any conditions outside of the approved design bases.

8. Confirmation 

Process

The confirmation process is commensurate with the specific- or 

general-licensee QA Program approved under 10 CFR Part 72, Subpart G, 

or 10 CFR Part 50, Appendix B, respectively. The QA Program ensures that 

the confirmation process includes provisions to ensure corrective actions are 

adequate and appropriate, have been completed, and are effective. The 

focus of the confirmation process is on the follow-up actions that must be 

taken to verify effective implementation of corrective actions. The measure 

of effectiveness is in terms of correcting the adverse condition and 

precluding repetition of significant conditions adverse to quality.

Procedures include provisions for timely evaluation of adverse conditions 

and implementation of any corrective actions required, including root cause 

evaluations and prevention of recurrence where appropriate. These 

procedures provide for tracking, coordinating, monitoring, reviewing, 

verifying, validating, and approving corrective actions, to ensure effective 

corrective actions are taken.

9. Administrative 

Controls

The administrative controls are in accordance with the specific- or 

general-licensee QA Program approved under 10 CFR Part 72, Subpart G, 

or 10 CFR Part 50, Appendix B, respectively. The QA Program ensures that 

the administrative controls include provisions that define:

• formal review and approval processes

• record retention requirements

• document control
B-28

Element Description

10. Operating 

Experience

The program references and evaluates applicable operating experience, 

both before renewal and will continue to do so as new operating experience 

is developed and made available after renewal, including:

• internal and industrywide condition reports

• internal and industrywide corrective action reports

• vendor-issued safety bulletins

• NRC Generic Communications

• applicable DOE or industry initiatives (e.g., HDRP)

• applicable research (e.g., Oak Ridge National Laboratory studies on 

bending responses of the fuel, Argonne National Laboratory and Central 

Research Institute of Electric Power Industry studies on hydride 

reorientation effects)

The review of operating experience clearly identifies any HBU fuel 

degradation as either age-related or event-driven, with proper justification for 

that assessment. Past operating experience supports the adequacy of the 

HBU Fuel Monitoring and Assessment Program.

Surrogate demonstration programs with storage conditions and fuel types 

similar to those in the licensed ISFSI/certified dry storage system that meet 

the guidance in Appendix D are a viable method to obtain operating 

experience.

New data/research on fuel performance from both domestic and 

international sources that are relevant to the licensed/certified HBU fuel in 

the dry storage system/ISFSI should be evaluated on a periodic basis and 

the AMP updated and revised as needed.
B-29

Element Description

References Center for Nuclear Waste Regulatory Analyses, “Extended Storage and 

Transportation: Evaluation of Drying Adequacy,” June 2013, NRC Contract 

No. NRC-02-07-006, 2013. ADAMS Accession No. ML13169A039.

Dominion Resources, Inc., “North Anna ISFSI, TN-32 High Burnup Dry 

Storage Research Project,” May 13, 2015. ADAMS Accession 

No. ML15133A082.

EPRI, “HBU Dry Storage Cask Research and Development Project, Final 

Test Plan,” February 27, 2014, DOE Contract No. DE-NE-0000593.

NRC, “Standard Review Plan for Spent Fuel Dry Storage Systems at a 

General License Facility,” NUREG-1536, Rev. 1, Washington, DC, 2010. 

ADAMS Accession No. ML101040620.

NRC, Interim Staff Guidance 11, “Cladding Considerations for the 

Transportation and Storage of Spent Fuel,” Revision 3, November 17, 2003. 

ADAMS Accession No. ML033230335.

a The AMP should provide a technical basis that supports the conclusion that the demonstration fuel is reasonably 

characteristic of the licensee’s stored fuel, in the event of differences in burnup, cladding composition (alloy 

types) and peak cladding temperatures observed during loading (See Appendix D). b The applicant will need to provide the expected upper bound moisture content based on its design-bases drying 

process. For example, if the design-bases drying process involves a vacuum drying method of evacuating a 

cask/canister to less than or equal to 3 torr and maintaining a constant pressure for 30 minutes after the 

cask/canister is isolated from the vacuum pump, the expected water content is about 0.43 gram-mole. (See 

NUREG-1536, Rev. 1.)

Table B-4. Formal Evaluations of Aggregate Information on HBU Fuel Performance

Year Assessment

1 Date—before HBU fuel 

exceeds the initial 

storage term a

Evaluate information obtained from a surrogate demonstration 

program for loading and initial period of storage along with other 

available sources of information. If surrogate demonstration program 

nondestructive examination (NDE) (i.e., cask/canister gas sampling, 

temperature data) data has not been obtained at this point, then the 

licensee has to provide evidence that the acceptance criteria in 

Element 6 are met or initiate a corrective action.

2 Date—10 years after 

Assessment 1 above

Evaluate, if available, information obtained from the destructive 

examination and NDE of the fuel placed into storage in a surrogate 

demonstration program along with other available sources of 

information. If the destructive examination data from a surrogate 

demonstration program has not been obtained at this point, then the 

licensee has to provide evidence to the NRC, by opening a 

cask/canister or separate effects surrogate experiments, that the 

acceptance criteria in Element 6 are met or initiate a corrective action.

3 Date—10 years after 

Assessment 2 above

Evaluate any new information.

a See Appendix F for a discussion of storage terms.

APPENDIX C

[RESERVED]

C-1

Upon consideration of public comments received on Draft NUREG-1927, Revision 1, the 

concepts regarding lead system inspections originally presented in Appendix C of the Draft 

NUREG-1927, Revision 1, have been incorporated into Chapter 3, “Aging Management 

Review,” of this Final NUREG-1927, Revision 1. Therefore, the content in Appendix C has been 

deleted, and Appendix C is reserved for future use.

APPENDIX D

SUPPLEMENTAL GUIDANCE FOR THE USE OF A DEMONSTRATION 

PROGRAM AS A SURVEILLANCE TOOL FOR CONFIRMATION OF 

INTEGRITY OF HIGH BURNUP FUEL DURING THE PERIOD OF 

EXTENDED OPERATION

D-1

Appendix D

Supplemental Guidance for the Use of a Demonstration Program as a 

Surveillance Tool for Confirmation of Integrity of High Burnup Fuel 

During the Period of Extended Operation

This guidance provides the U.S. Nuclear Regulatory Commission (NRC) staff a basis for 

reviewing if a demonstration of high burnup (HBU) fuel has the necessary properties to qualify 

as one method that an applicant might use in license and certificate of compliance renewal

applications to confirm the integrity of HBU fuel during continued storage.

D.1 Discussion

The experimental confirmatory basis that low burnup fuel (≤45 GWd/MTU) will maintain its 

integrity in dry cask storage over extended time periods was provided in a demonstration test 

(NRC, 2003; Bare, et al., 2001; Einziger, et al., 2003). A similar confirmation test, which 

includes information over a similar length of the time available for low burnup fuel, does not exist 

for other light water reactor (LWR) fuels, HBU fuel 1 and mixed oxide fuels. Certification and 

licensing HBU fuel for storage was permitted for an initial 20-year-term using the guidance 

contained in Interim Staff Guidance (ISG) 11 (NRC 2003), which was based on short-term 

laboratory tests and analysis that may not be applicable to the storage of HBU fuel beyond 

20 years, particularly with the current state of knowledge regarding HBU fuel cladding 

properties.

One concern stated in ISG-11 was the potential detrimental effects, such as reduced ductility, of 

hydride reorientation on cladding behavior (NRC, 2003). Research performed in Japan and the 

United States indicated that: (1) hydrides could reorient at a significantly lower stress than 

previously believed (Billone, et al., 2013; Kamimura, 2010; Daum, et al, 2006), and (2) the radial 

hydrides could raise the cladding ductile-to-brittle transition temperature enough to compromise 

the ability of the cladding to withstand stress without undergoing brittle failure (Billone,

et al., 2013). This phenomenon could influence the retrievability of HBU fuel assemblies and 

result in operational safety concerns in the handling of individual assemblies as HBU fuel

cooled. Circumferential zirconium hydrides in the fuel cladding regions would dissolve into the 

fuel cladding during drying and reprecipitate (reorient) as radial hydrides as the fuel cladding 

cooled. Thus, fuel cladding with radial hydrides that is below a ductile-to-brittle transition 

temperature could be too brittle to retrieve (remove from the DSS) on an assembly basis. The 

maximum temperatures and internal rod pressures in ISG-11 were recommended to mitigate 

hydride reorientation and are applicable to HBU fuel during the initial 20-year storage, as the 

decay heat of HBU fuel is expected to maintain cladding temperatures above a ductile-to-brittle 

transition temperature (about 200 degrees Celsius, or about 392 degrees Fahrenheit).

There is no evidence to suggest that HBU fuel cannot similarly be stored safely and then 

retrieved for time periods beyond 20 years, but the supporting experimental data is not 

extensive. Therefore, confirmatory data or a commitment to obtain data on HBU fuel and taking 

appropriate steps in an aging management plan (AMP) will provide further information that will 

be useful in evaluating the safe handling of individual assemblies of HBU fuel for 

extended durations.

 1 High burnup fuel is fuel with burnup ≥ 45 GWd/MTU.
D-2

A demonstration program could provide an acceptable method for an applicant to demonstrate 

compliance with the cited regulations for storage of light water reactor fuels (LWR) for periods of 

greater than 20 years by:

• confirming the expected fuel conditions, based on technical arguments made in ISG-11

(NRC, 2003), after a substantial storage period that is sufficiently long (about 10 years) 

to extrapolate the findings to the storage duration of interest (The behavior of the 

cladding for the period of extended operation will depend on its physical condition at the 

end of the initial 20-year storage period.)

• providing data for benchmarking, confirming predictive models, and updating aging 

management plans

• confirming the time-limited aging analysis (TLAA) cladding creep predictions that are 

the basis for the guidance recommendation for the maximum temperature in ISG-11

(NRC, 2003) are not exceeded and that sufficient creep margin exists for the extended 

storage period

• determining the system is sufficiently dry to eliminate moisture-driven degradation 

from consideration

• providing operating experience on the fuel behavior and drying procedure as input to an 

AMP on the behavior of the fuel

• identifying any aging effects that may be missed through short-term accelerated studies 

and analyses

Monitoring of the fuel temperatures, gas composition, and other conditions in the canister or 

cask combined with physical examination of the fuel at periodic intervals should be able to 

provide confirmation if:

• The models of the phenomena used for the first 20-year predictions can be used for the 

TLAA beyond 20 years.

• The condition of the fuel, after an appropriately long period of storage, does not degrade.

• New degradation mechanisms are not being exhibited.

Extrapolation outside the recorded data carries risk, but that risk can be minimized if the length 

of the extrapolation is reduced and those extrapolations are updated as the demonstration 

continues to monitor and measure fuel properties.

D.2 Technical Review Guidance

The applicant may use the results of a completed demonstration or an ongoing demonstration 

if the conditions of the demonstration meet the requirements stated below for the fuels and 

conditions of storage for which the term is to be renewed. The approach in this guidance can 

be applied to a generic demonstration program or a site/system-specific program as long as 

the demonstration’s parameters are reasonably applicable to the applicant’s fuel type and 

characteristics.
D-3

The technical reviewers should establish that the following conditions are met if the 

demonstration is to be used by the applicant to support fuel assembly conditions for storage of 

HBU fuel beyond 20 years and to be applicable to support a license or certificate application:

(1) That the maximum burnup of the fuel in the application is less than the burnup of the fuel 

in the demonstration. If the burnup is higher than that in the demonstration, the 

applicant should provide evidence, based on characteristics of the fuel, derived either 

from reactor rod qualification testing or other separate effects tests, that the 

demonstration fuel is reasonably characteristic of the stored fuel and the added burnup 

will not change the results determined by the demonstration. Similarly, if there is a 

different cladding type used, arguments based on comparison of composition and 

fabrication technique (e.g., stress-relieved and annealed, recrystallized) should justify 

the use of the demonstration results.

(2) If the applicant uses direct observations of the rod behavior to imply the condition of 

the rods in its system, either (1) the temperatures in the demonstration must bound the 

temperatures in the application, or (2) if the applicant uses predictive tools that have 

been confirmed by the demonstration, then the temperatures of the rods in the 

application do not have to be bounded by the temperature of the rods in the 

demonstration. The temperature models used in the application should either be 

benchmarked (1) against the demonstration temperature data, or (2) against actual 

measured rod temperature data in the same temperature range.

(3) If the applicant is using gas analysis or another gas detection method to establish the 

condition of the fuel, then the interior of a demonstration canister or cask should be 

quantitatively monitored for, at a minimum, moisture, oxygen, and fission gas. The 

duration and frequency of the gas monitoring should be determined by analysis of the 

potential degradation. Gases should always be quantitatively monitored before opening 

of the canister. If the applicant claims that no galvanic degradation is feasible, then, if 

after drying, moisture is detected in the canister, moisture and hydrogen should be 

monitored at a reasonable frequency to be determined by the applicant until the moisture 

disappears. Gas monitoring is not expected during movement of the canister. If the 

applicant is using the gas analysis to show no breaches would occur during transport, 

gas quantitative monitoring must be conducted before and after transport.

(4) Temperature monitoring should be conducted at a frequency that is suitable 

for determining the profile over the duration of the demonstration.

(5) If possible, some population of stored rods should be examined whenever the system is 

opened. These rods should be extracted from the fuel assembly to determine properties 

of the rods that affect degradation such as cladding creep, fission gas release, hydride 

reorientation, cladding oxidation, and mechanical properties.

(6) The demonstration program fuel shall include at least two full fuel assemblies. The 

assemblies may be reconstituted.

(7) Data from the demonstration program must be indicative of a storage duration long 

enough to justify extrapolation to the total storage time requested but no less than 
D-4

10 years if the data is to be used to support license extension from the initial 20 years 

to an additional 40 years.

2

D.3 References

Bare, W.C, L.D. Torgerson, “Dry Cask Storage Characterization Project—Phase 1: 

CASTOR V/21 Cask Opening and Examination,” NUREG/CR-6745, Idaho National Engineering 

and Environmental Laboratory, Idaho Falls, ID, 2001. Agencywide Documents Access and 

Management System (ADAMS) Accession No. ML013020363.

Billone, M.C., T.A. Burtseva, R.E. Einziger, “Ductile-to-Brittle Transition Temperature for 

High-Burnup Cladding Alloys Exposed to Simulated Drying-Storage Conditions,” Journal 

of Nuclear Materials, Volume 433, Issues 1–3, pages 431–448, February 2013.

Daum, R.S., et al., 2006. “Radial-Hydride Embrittlement of High-burnup Zircaloy-4 Cladding”, 

Journal of Nuclear Science and Technology, Vol. 43, No. 9, p.1054, 2006.

Einziger, R.E., et al., “Examination of Spent PWR Fuel Rods after 15 Years in Dry Storage,”

NUREG/CR-6831, Argonne National Laboratory, Argonne, IL, 2003. ADAMS Accession 

No. ML032731021.

Kamimura, K. “Integrity Criteria of Spent Fuel for Dry Storage in Japan,” Proceedings of 

the International Seminar on Interim Storage of Spent Fuel, ISSF 2010, page VI3-1, Tokyo 

November 2010.

U.S. Nuclear Regulatory Commission (NRC), Interim Staff Guidance 11, “Cladding 

Considerations for the Transportation and Storage of Spent Fuel,” Rev. 3, November 17, 2003.

NRC, “The Use of a Demonstration Program as a Surveillance Tool for Confirmation of Integrity 

for Continued Storage of HBU Fuel Beyond 20 Years,” ISG-24, Revision 0. Washington, DC, 

2014. ADAMS Accession No. ML14058B166.

 2 A demonstration is to provide that there was satisfactory performance during the first 20 years and that the 

results could be extrapolated to support an additional 40 years. The NRC staff agreed that a demonstration 

of ≤ 10 years storage duration is insufficient to support these goals.
APPENDIX E

CONSIDERATIONS FOR RENEWALS OF CERTIFICATES 

OF COMPLIANCE

E-1

Appendix E

Considerations for Renewals of Certificates of Compliance

E.1 Development of Time-Limited Aging Analyses and Aging Management Programs

From Title 10 of the Code of Federal Regulations (10 CFR) 72.240, “Conditions for Spent Fuel 

Storage Cask Renewal,” the certificate of compliance (CoC) renewal application must include 

time-limited aging analyses (TLAAs), if applicable, and aging management programs (AMPs). 

The CoC holder, as the applicant for the CoC renewal and the owner of the storage cask or dry 

storage system (DSS) design, will develop the TLAAs and AMPs. The renewal application 

should address age-related degradation of the DSS design in a bounding manner (i.e., fully 

address pertinent aging mechanisms and effects in all possible service environments where the 

DSS is being used and can be used). If there is an AMP that may not be applicable to certain 

general licensees, because of the service environment in which the DSS is located, the AMP 

should specify this. In addition, the CoC holder and the staff should consider the need for a 

CoC condition to specify this potential limited use of the AMP, so it is clear for general 

licensee implementation.

For a CoC renewal that encompasses CoC amendments with different design bases, the CoC 

holder will need to address in the renewal application how the TLAAs or AMPs apply to each 

amendment covered by the CoC. For example, if different materials are used or different SSCs 

are part of the DSS design in different CoC amendments, or if different environments 

(e.g., underground vs. aboveground system) are reflected in different CoC amendments, 

then there may be different TLAAs or AMPs specified for the individual amendments.

E.2 Implementation of AMPs

In approving the renewal of the DSS design, the U.S. Nuclear Regulatory Commission (NRC)

may revise the CoC to include terms, conditions, and specifications that will ensure the safe 

operation of the DSS during the period of extended operation, including but not limited to, terms, 

conditions, and specifications that will require the implementation of an AMP by a general 

licensee, in accordance with 10 CFR 72.240(e). General licensees’ implementation of AMPs 

is subject to NRC inspection.

Regulations in 10 CFR 72.212, “Conditions of General License Issued under § 72.210,” provide

requirements for general licensees using approved CoCs. Regulations in 10 CFR 72.212(b)(11) 

require general licensees to comply with the terms, conditions, and specifications of the CoC, 

including but not limited to, the requirements of any AMP put into effect as a condition of the 

NRC approval of a CoC renewal application in accordance with 10 CFR 72.240.

General licensees (CoC users) are responsible for implementing the AMPs. To document 

the licensee’s compliance with the renewed CoC, a future or existing general licensee should 

include in its 10 CFR 72.212(b)(5) evaluation (including the results of the review and 

determination per 72.212(b)(6) and 72.212(b)(8)) how it will meet the new CoC terms, 

conditions, or specifications for aging management. Note that the renewed CoC may include a 

condition for the general licensee to do so. As part of this evaluation, general licensees should 

consider any granted exemptions or 10 CFR 72.48 changes that could affect aging 

management.
E-2

The general licensee should update the 10 CFR 72.212(b)(5) evaluation before entering the 

period of extended operation. Considering timely renewal provisions (See Section 1.4.5), 

update of the 10 CFR 72.212(b)(5) evaluation before the loaded systems enter the period of 

extended operation may not be possible. In such cases, the reviewer should ensure that timing 

for update of the 10 CFR 72.212(b)(5) evaluation is addressed in the application in a clear 

manner. Also, any CoC condition related to the update of the 10 CFR 72.212(b)(5) evaluation 

may include timing provisions.

The AMP-related information in the DSS FSAR will be implemented by general licensees 

as-written. If a general licensee wishes to deviate from the DSS FSAR, it must evaluate any 

such deviation under the provisions of 10 CFR 72.48. If AMP details in the FSAR specify that 

the AMP is not applicable for certain users (e.g., if it is not applicable in certain climates or 

environments), the general licensee can include the technical justification in its 10 CFR 72.212 

report for not implementing such an AMP. The general licensee would need to evaluate any 

changes to the 10 CFR 72.212 report or any deviations from the DSS FSAR, using the 

requirements of 10 CFR 72.48 (See NRC Regulatory Issue Summary 2012-05, “Clarifying 

the Relationship Between 10 CFR 72.212 and 10 CFR 72.48 Evaluations”).

E.3 Corrective Actions

As discussed in Section 3.6.1.7, corrective actions are measures to be taken when the AMP 

acceptance criteria are not met. Corrective actions are critical for maintaining the intended 

functions of the structures, systems, and components during the initial storage term as well as 

the period of extended operation. The CoC holder should discuss in its renewal application any 

applicable and appropriate corrective actions to be taken if the AMP acceptance criteria are not 

met.

A general licensee will use its Corrective Action Program (CAP) (that is consistent with the 

criteria in 10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants 

and Fuel Reprocessing Plants”) to capture and address aging effects identified in the period of 

extended operation. At a minimum, all conditions that do not meet the AMP acceptance criteria 

should be entered into the CAP. The general licensee’s CAP should be able to respond to and 

adequately address and rectify any ISFSI or DSS aging issues. Also, the CAP’s response to 

address any DSS aging issues should include any specific corrective actions specified in the 

CoC renewal application.
APPENDIX F

STORAGE TERMS

F-1

Appendix F

Storage Terms

F.1 Introduction

This appendix provides a flow chart for calculating storage terms of a dry storage system (DSS)

or a DSS structure, system, and component (SSC) loaded during either the initial storage period 

or renewal period(s) of a certificate of compliance (CoC).

F.2 Storage Term Defined

The storage term (length of time a DSS can remain loaded) is determined by the period 

specified in the applicable CoC in effect at the time the DSS is placed into service (from Title 10 

of the Code of Federal Regulations (10 CFR) 72.212(a)(3) and Ref. 1). The storage period 

begins when the DSS is first used by the general licensee to store spent fuel 

(10 CFR 72.212(a)(3)).1

 The clock starts when the loaded cask has been deployed in the 

independent spent fuel storage installation (76 FR 8872). If a CoC is not renewed, upon 

expiration, casks loaded under that CoC would need to be removed from service when they 

reach the end of their storage term. An AMP for a renewed CoC commences at the end of the 

initial storage period for each loaded DSS (see Section 3.6.2).

• If the DSS is loaded during the initial CoC term (e.g., 20 years) and the CoC is not 

renewed, the storage term is the entirety of the initial CoC term (e.g., 20 years).

• If the DSS is loaded during the initial CoC term and the CoC is renewed once, the 

storage term is the remaining time in the initial CoC term added to the entirety of the 

first renewal period.

• If the DSS is loaded during the first renewal period (e.g., 40 years), and the CoC is 

not subsequently renewed, the storage term is the entirety of the first renewal period 

(e.g., 40 years).

• If the DSS is loaded during the first renewal period, and the CoC is subsequently 

renewed, the storage term is the remaining time in the first renewal period added to 

the entirety of the subsequent renewal period (cumulative).

F.3 Flowchart for Calculating Storage Terms

A flowchart is provided below to assist the user in calculating the storage term for a DSS loaded 

under a CoC.

 1 The storage period for a particular DSS SSC begins when it is first used to store spent fuel, regardless if the 

SSC is later stored at a different location.
F-2

Loaded 

during initial 

CoC term?

Subsequently 

renewed?

Start

Renewed?

A = 0

B= full length of first 

renewal period 

(years)

C = 0

A = 0

B = remaining time in first 

renewal period (years)

C = any subsequent 

renewal period(s) 

[cumulative, in years]

A = remaining time in the 

initial CoC term (years)

B = full length of first 

renewal period (years)

C = any subsequent 

renewal period(s) 

[cumulative, in years]

A = full length of initial 

CoC term (years)

B = 0

C = 0

Yes

No

Yes

No

No

Yes

Yes

Loaded during 

first renewal 

period?

Length of storage term = A + B + C

Figure F-1. Flowchart for calculating storage terms

F.4 References

U.S. Nuclear Regulatory Commission, “10 CFR Part 72, License and Certificate of Compliance 

Terms, Final Rule,” Federal Register, Vol. 76, No. 32, February 16, 2011, pp. 8872–8892 

(76 FR 8872).






NUREG-1927, Rev. 1

Final

Standard Review Plan for Renewal of Specific Licenses and Certificates of 

Compliance for Dry Storage of Spent Nuclear Fuel