ML24121A004

DG-3058 (RG 3.78 Rev 0) ACRS Pre-Decisional Version

ML24121A004
Person / Time
Issue date:	04/30/2024
From:	Darrell Dunn NRC/NMSS/DFM/IOB
To:
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RG 3.78 Rev 0 DG-3058
Download: ML24121A004 (15)

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U.S. NUCLEAR REGULATORY COMMISSION DRAFT REGULATORY GUIDE DG-3058 Proposed new Regulatory Guide 3.78 Issue Date: Month 2024 Technical Lead: Darrell Dunn Pre-Decisional for ACRS Consideration This RG is being issued in draft form to involve the public in the development of regulatory guidance in this area. It has not received final staff review or approval and does not represent an NRC final staff position. Public comments are being solicited on this DG and its associated regulatory analysis. Comments should be accompanied by appropriate supporting data. Comments may be submitted through the Federal rulemaking Web site, http://www.regulations.gov, by searching for draft regulatory guide DG-XXXX. Alternatively, comments may be submitted to the Office of Administration, Mailstop: TWFN 7A-06M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN:

Program Management, Announcements and Editing Staff. Comments must be submitted by the date indicated in the Federal Register notice.

Electronic copies of this DG, previous versions of DGs, and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/. The DG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. MLXXXXXXXXX. The regulatory analysis may be found in ADAMS under Accession No. MLXXXXXXXXX.

Pre-Decisional for ACRS Consideration ACCEPTABLE ASME SECTION XI INSERVICE INSPECTION CODE CASES FOR 10 CFR PART 72 A. INTRODUCTION Purpose This regulatory guide (RG) describes methods and procedures acceptable to the U.S. Nuclear Regulatory Commission (NRC) staff that were developed by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI, Inservice Inspection (Ref. 1),

Code Cases. The NRC staff NRC has determined these Code Cases to be acceptable for use by specific licensees for independent spent fuel storage installations (ISFSIs), general licensees, and certificate of compliance (CoC) holders licensed under 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 (Ref. 2).

Applicability This RG applies to applicants for renewal of CoCs and specific ISFSI licenses, and to holders of CoCs and specific and general licenses subject to the regulatory requirements for spent fuel storage renewals in 10 CFR Part 72.

Applicable 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, contains requirements, procedures, and criteria for the issuance of licenses to receive, transfer, and possess power reactor spent fuel, power-reactor-related greater-than-Class-C waste, and other radioactive materials associated with spent fuel storage in an ISFSI, as well as the terms and conditions under which the NRC will issue these licenses. The regulations in this part also establish requirements, procedures, and criteria for the issuance of CoCs approving spent fuel storage cask designs.

o 10 CFR 72.42, Duration of license; renewal, provides requirements for the duration of ISFSI specific licenses and for applications for license renewal, including requirements for time-limited aging analyses (TLAAs) and aging management programs (AMPs).

Pre-Decisional for ACRS Consideration DG-3058 Page 2 Pre-Decisional for ACRS Consideration o

10 CFR 72.240, Conditions for spent fuel storage cask renewal, provides requirements for applications for the renewal of CoCs for spent fuel storage cask designs, including requirements for TLAAs and AMPs.

Related Guidance NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel (Ref. 3), provides guidance for the NRCs safety review of renewal applications for ISFSI specific licenses and CoCs for spent fuel storage cask designs.

NUREG-2214, Managing Aging Processes In Storage (MAPS) Report (Ref. 4), provides a generic technical basis for renewal of ISFSI specific licenses and CoCs for spent fuel storage cask designs.

Purpose of Regulatory Guides The NRC issues RGs to describe methods that are acceptable to the staff for implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific issues or postulated events, and to describe information that the staff needs in its review of applications for permits and licenses. Regulatory guides are not NRC regulations and compliance with them is not required.

Methods and solutions that differ from those set forth in RGs are acceptable if supported by a basis for the issuance or continuance of a permit or license by the Commission.

Paperwork Reduction Act This RG provides voluntary guidance for implementing the mandatory information collections in 10 CFR Part 72 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved by the Office of Management and Budget (OMB), under control number 3150-0132. Send comments regarding this information collection to the FOIA, Library, and Information Collections Branch (T6-A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by email to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0132), Office of Management and Budget, Washington, DC, 20503.

This RG provides voluntary guidance for implementing the mandatory information collections in 10 CFR Part 72 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et. seq.). These information collections were approved by the Office of Management and Budget (OMB), approval numbers 3150-0011 and 3150-0151. Send comments regarding this information collection to the Information Services Branch (T6-A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-0011 and 3150-0151), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street, NW Washington, DC20503; e-mail:

oira_submission@omb.eop.gov.

Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number.

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Pre-Decisional for ACRS Consideration DG-3058 Page 4 Pre-Decisional for ACRS Consideration B. DISCUSSION Reason for Issuance This RG provides guidance to specific licensees for independent spent fuel storage installations (ISFSIs), general licensees, and certificate of compliance (CoC) holders licensed under Title 10 of the Code of Federal Regulations (10 CFR) Part 72, by identifying ASME Section XI Code cases that the NRC staff has determined to be acceptable for use. This guidance is applicable to codes and standards for inservice inspection of confinement boundary components and aging management activities associated with the renewals of ISFSI licenses and CoCs for spent fuel storage systems.

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Background===

The NRCs regulatory requirements for ISFSI and storage system CoC renewals are contained in 10 CFR Part 72. In addition to the regulations, the NRC has developed implementing guidance for specific ISFSI license and CoC renewals. A summary of the regulatory requirements, guidance, and other applicable documents for specific ISFSI license and storage system CoC renewals related to this RG are described in the following paragraphs.

10 CFR Part 72 Regulatory Requirements The regulations in 10 CFR Part 72 include the requirements to obtain and renew (1) a specific ISFSI license, and (2) a storage system CoC. In addition, the regulations in 10 CFR Part 72 include the requirements for the use of a storage system CoC by a general licensee. The safety review conducted for a specific license or CoC is primarily based on the information the applicant provides in a Safety Analysis Report (SAR) to show that the design and operation meet the appropriate requirements in 10 CFR Part 72.

For the renewal of a specific ISFSI license, 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 users 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.

Both the specific-license ISFSI and the CoC renewal applications must contain requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI or dry storage system (DSS) that address aging mechanisms and aging effects that could affect structures, systems, and components relied upon for the safe storage of spent fuel. General licensees incorporate the CoC renewal requirements via 10 CFR 72.212, Conditions of general license issued under § 72.210.

Renewal applications must include (1) TLAAs, 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) AMPs for management of issues associated with aging that could adversely affect structures, systems, and components important to safety.

NRC Guidance NRC guidance on 10 CFR Part 72 renewals is contained in NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel. NUREG-1927 provides guidance for the general information, scoping evaluation information, and aging management information that should be included in a renewal application. The guidance provides information on TLAAs and AMPs, including learning AMPs that consider and respond to operating

Pre-Decisional for ACRS Consideration DG-3058 Page 5 Pre-Decisional for ACRS Consideration experience. It also provides guidance on considerations for CoC renewals and the general license framework, including guidance on general licensees implementation of AMPs.

NUREG-1927, Section 3.6 provides detailed guidance on AMPs for renewal applications. The purpose of an AMP is to monitor and control the degradation of SSCs within the scope of renewal so that aging effects will not result in loss of intended functions during the period of extended operation. As noted in NUREG-1927, 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.

The NRC guidance in NUREG-1927 was augmented by NUREG-2214, Managing Aging Processes In Storage (MAPS) Report. The MAPS Report evaluates known aging degradation mechanisms to determine if they could affect the ability of dry storage system components to fulfill their safety functions in the 20- to 60-year period of extended operation. The MAPS Report also provides examples of aging management programs that are considered generically acceptable to address the credible aging mechanisms to ensure that the design bases of dry storage systems will be maintained. An applicant for a renewed license or certificate of compliance may reference the information in the MAPS Report to support its aging management review and proposed aging management programs.

ASME Code Cases Provisions of the ASME Code have been used since 1971 as one part of the framework to establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety in nuclear power applications. A broad spectrum of stakeholders participate in the consensus codes and standards development process, including the development of Code Cases and revisions to the ASME B&PV Code. This broad participation helps to ensure that the various stakeholder interests are considered.

This RG is similar to other RGs for power reactors (NRC Regulatory Guides for Division 1) that address acceptable and unacceptable ASME Section III Code Cases including RG 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III (Ref. 5), and RG 1.193, ASME Code Cases Not Approved for Use (Ref. 63). In addition, RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 (Ref. 76), lists the acceptable and conditionally acceptable ASME Section XI Code Cases applicable to NRC licenses under 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 87). In particular, this RG is similar to RG 1.147 but is specifically focused on ASME Section XI code cases applicable to 10 CFR Part 72 licensees which are not addressed in the RGs listed above.

Pre-Decisional for ACRS Consideration DG-3058 Page 6 Pre-Decisional for ACRS Consideration ASME Code Case N-860 ASME Code Case (CC) N-860, Examination Requirements and Evaluation Standards for Spent Nuclear Fuel Storage and Transportation Containment System, was approved by the ASME board of Nuclear Codes and Standards (BNCS) in July 2020 (Ref. 98). ASME CC N-860 presumes that chloride-induced stress corrosion cracking (CISCC) is the most credible and bounding degradation mechanism that might challenge the confinement integrity of SNF storage and transportation containment systems, based on current knowledge and operating experience. ASME CC N-860 is a Section XI, Division 1, Code Case that provides rules for in-service inspection of dry storage canisters, standardizes the methodology for the number of systems inspected, the basis for which systems are selected for inspection, the initial frequency of inspection, the methodology for increasing inspection frequency if aging effects are found, and the criteria that have to be met in order to decrease inspection frequency.

ASME CC N-860 is different than traditional ASME Code Cases that typically provide alternatives to the existing code requirements. Instead, CC N-860 provides the code requirements for ISI of canisters that are not provided elsewhere in the ASME B&PV Code.

ASME Code Case N-860, Subarticle-2700 provides the requirements for inspection intervals and inspection populations following both the initial inspection after entering the storage period of operation (SPEO) and subsequent inspections at the specified intervals. The requirements can be modified based on the susceptibility of ISFSI sites to CISCC, which is determined using the methodology described in Electrical Power Research Institute (EPRI), Susceptibility Assessment Criteria for Chloride-Induced Stress Corrosion Cracking (CISCC) of Welded Stainless Steel Canisters for Dry Cask Storage Systems, EPRI-3002005371 (Ref. 109). For ISFSI sites with a CISCC susceptibility ranking of 7 and below, increases in the inspection intervals and decreases in populations are allowed depending on the results of the Screening Examination described in ASME CC N-860, Subarticle-2200, with the stipulation that the maximum inspection interval is 20 years and the minimum inspection population is one canister per ISFSI site. However, Subarticle-2700 does not allow increases in the inspection interval or decreases in the inspection population for ISFSI sites which have a ranking of 8 and above.

Many specifically licensed ISFSIs and CoCs have gone through renewal prior to the development of CC N-860 and thus have NRC approved AMPs. For ISFSIs and CoCs that use welded stainless steel canister based systems, the NRC approved AMPs are based on inspections for identification of aging effects and include specific details on inspection methods, personnel qualifications, inspection frequency and acceptance criteria which overlap with the rules in CC N-860. These approved AMPs provide reasonable assurance of adequate protection for the important to safety structures systems and components and the use of CC N-860 is not required to meet the requirements in 10 CFR 72.42 and 10 CFR 72.240.

Nonetheless, the NRC staff believes that adoption of CC N-860 could be beneficial to NRC licensees, CoC holders, and industry as a whole by standardizing inspection requirements and the technical bases for canister selection and inspection frequency. The NRC licensees and CoC holders could use the evaluation criteria in 10 CFR 72.48 to determine if CC N-860 could be adopted without an amendment. In addition, applicants for renewed ISFSIs licenses and CoCs could use ASME CC N-860 in the applications.

NRC RG 3.76 and NEI 14-03 Revision 2 NRC RG 3.76, Implementation of Aging Management Requirements for Spent Fuel Storage Renewals (Ref. 110), endorses, with clarifications, NEI 14-03, Revision 2, Format, Content and Implementation Guidance for Dry Cask Storage Operations-Based Aging Management (Ref. 121). NEI

Pre-Decisional for ACRS Consideration DG-3058 Page 7 Pre-Decisional for ACRS Consideration 14-03 provides an operations-based, learning approach to aging management for the storage of spent fuel, which builds on the lessons learned from industrys experience with aging management for reactors.

Specifically, NEI 14-03 provides a framework for sharing operating experience through an industry-developed database called the ISFSI Aging Management Institute of Nuclear Power Operations Database (AMID). NEI 14-03 also includes a framework for learning AMPs through the use of tollgates, which offer a structured approach for periodically assessing operating experience and data from applicable research and industry initiatives. The AMID database provides operating experience information and a basis to support licensees future changes to AMPs. The AMID database and tollgates are considered key elements in ensuring the effectiveness of aging management activities and the continued safe storage of spent fuel during the period of extended operation (PEO).

ISFSI Operating Experience and Examination Results To fulfill the requirements of AMPs included in renewed ISFSI licenses and dry storage system CoCs, examinations of welded austenitic stainless steel dry storage system canisters have been conducted at ISFSI sites with a range of CISCC susceptibility rankings. No evidence of CISCC or localized corrosion, such as pitting corrosion or crevice corrosion, have been identified during examinations of welded austenitic stainless steel canisters at any site conducted to date.

For most ISFSI sites, the lack of any observed indications for localized corrosion or CISCC is consistent with expectations because the majority of operating ISFSI sites are not located in close proximity to a source of chloride salts such as a marine coastline, a cooling tower, or a roadway that is treated with deicing salts. In addition, these sites are typically characterized as having low or average absolute humidity values. These sites of low CISCC susceptibility have a CISCC susceptibility ranking of less than 4.

In addition to examinations of welded austenitic stainless steel dry storage system canisters, deposits on the canisters have been collected and analyzed. The collection methodology and analysis results of the collected samples are publicly available in the following Sandia National Laboratories (SNL) reports via the U.S. Department of Energy (DOE) Office of Scientific and Technical Information web site (www.osti.gov):

SAND2020-13674, Analysis of Dust Samples Collected from an Inland ISFSI Site (Site A),

December 2020 (Ref. 132).

SAND2020-14144, Analysis of Dust Samples Collected from an Inland ISFSI Site (Site B),

December 2020 (Ref. 143).

SAND2022-10884, Analysis of Dust Samples Collected from a Near-Marine East Coast ISFSI Site (Site C), August 2022 (Ref. 154).

The evaluations concluded that the risk of CISCC at these sites is low based on the minimal surface concentrations of chloride salts. In addition, when combined with relatively high concentrations of nitrate (that acts as a corrosion inhibitor) at the sites, the risk of canister cracking from CISCC may be very low.

Conditions for CISCC A combination of a susceptible material, sufficient applied or residual tensile stresses, and an environment where chloride ions are present is necessary to induce CISCC. NRC staff determined that CISCC was a potential aging mechanism for dry storage system designs that use welded austenitic stainless steel canisters and which may be exposed to a range of environments. Nuclear industry operational experience with CISCC of nuclear power plant piping systems and tanks was documented in

Pre-Decisional for ACRS Consideration DG-3058 Page 8 Pre-Decisional for ACRS Consideration NRC Information Notice 2012-20, Potential for Chloride-Induced Stress Corrosion Cracking of Austenitic Stainless Steel and Maintenance of Dry Cask Storage Systems, issued November 14, 2012 (Ref. 165).

The CISCC operational experience documented in NRC Information Notice 2012-20 is limited to events where austenitic stainless steel components were exposed to marine shoreline atmospheric conditions. While the NRC staff considered CISCC of welded austenitic stainless steel canisters to be a credible aging mechanism in sheltered environments, the staff recognized that not all ISFSI sites would have the required combination of chloride containing salts that could be deposited on passively cooled canisters and sufficient humidity necessary for deliquescence of the deposited chloride salt to form an aqueous solution with chloride ions.

A wide variety of compounds may be deposited on the surface of a passively cooled storage canister. Non-chloride containing salts such as nitrate and or sulfate containing species are common and may also be deposited on the canisters or may form by the chemical transformation of deposited sea salt aerosols. However, Chi, et al., showed that as sea salt aerosols age, the aerosols are chemically transformed from a chloride rich composition to a combination of sodium sulfate and sodium nitrate with the original chloride completely removed (Ref. 176). In addition, nitrate and sulfate are known to inhibit localized corrosion of stainless steels in chloride environments (Ref. 187-210). Previous testing conducted in high temperature, concentrated, magnesium chloride (MgCl2) solutions has also shown that sodium nitrate (NaNO3) is an effective inhibitor for chloride stress corrosion cracking of stainless steels (Ref. 221-243). These factors limit the effects of the deposits on the canisters.

The available information for susceptibility of canisters includes the characterization of salts deposited on canisters at a range of ISFSI sites, the known evolution of sea sat deposits with time, extensive CISCC testing of austenitic stainless steels, and ISFSI operating experience. The sum of this information conclusively shows that CISCC of welded austenitic stainless steel canisters at low ranked ISFSI sites is very unlikely.

Consideration of International Standards The International Atomic Energy Agency (IAEA) works with member states and other partners to promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops Safety Requirements and Safety Guides for protecting people and the environment from harmful effects of ionizing radiation. This system of safety fundamentals, safety requirements, safety guides, and other relevant reports, reflects an international perspective on what constitutes a high level of safety. To inform its development of this RG, the NRC considered IAEA Safety Requirements and Safety Guides pursuant to the Commissions International Policy Statement (Ref. 25) and Management Directive and Handbook 6.6, Regulatory Guides (Ref. 26).

Harmonization with International Standards The NRC has a goal of harmonizing its regulatory guidance with documents issued by the International Atomic Energy Agency (IAEA) to the extent practical. Although the NRC does not endorse the following IAEA safety standard(s) and/or guide(s), this RG incorporates similar guidelines and is consistent with the basic safety principles provided in them.

Specific Safety Requirements No. SSR-4: Safety of Nuclear Fuel Cycle Facilities (Ref. 274) includes the requirements address implementation of an ageing management program to manage the ageing of items important to safety so that the required safety functions are fulfilled over the entire operating lifetime of the nuclear fuel cycle facility.

Pre-Decisional for ACRS Consideration DG-3058 Page 9 Pre-Decisional for ACRS Consideration Specific Safety Guide No. SSG-15 Storage of Spent Nuclear Fuel (Ref. 285) includes guidance for the SSG-4 requirements on implementation of an ageing management programs for items important to safety.

Documents Discussed in Staff Regulatory Guidance This RG endorses the use of ASME Code Cases applicable to NRC licensees under 10 CFR Part

72. The ASME Code Cases may contain references to other codes, standards or third-party guidance documents (secondary references). If a secondary reference has itself been incorporated by reference into NRC regulations as a requirement, then licensees and applicants must comply with that standard as set forth in the regulation. If the secondary reference has been endorsed in a RG as an acceptable approach for meeting an NRC requirement, then the standard constitutes a method acceptable to the NRC staff for meeting that regulatory requirement as described in the specific RG. If the secondary reference has neither been incorporated by reference into NRC regulations nor endorsed in a RG, then the secondary reference is neither a legally-binding requirement nor a generic NRC approved acceptable approach for meeting an NRC requirement. However, licensees and applicants may consider and use the information in the secondary reference, if appropriately justified, consistent with current regulatory practice, and consistent with applicable NRC requirements.

Pre-Decisional for ACRS Consideration DG-3058 Page 10 Pre-Decisional for ACRS Consideration C. STAFF REGULATORY GUIDANCE The NRC staff endorses the ASME Code Cases in Table 1 as generally acceptable for use in complying with the requirements in 10 CFR 72.42 and 10 CFR 72.240.

Table 1. Acceptable Section XI Code Cases for 10 CFR Part 72 Licensees and CoC holders Code Case Number Code Case Title Date N-860 Inspection Requirements and Evaluation Standards for Spent Nuclear Fuel Storage and Transportation Containment SystemsSection XI, Division 1;Section XI, Division 2 7/6/2020 ASME Code Case N-860 The NRC staff endorses ASME CC N-860 as generally acceptable for use in complying with the requirements in 10 CFR 72.42 and 10 CFR 72.240.

The NRC staff provides an additional alternative to ASME CC N-860, Subarticle-2700, Evaluate Inspection Interval and Inspection Populations, Subsubarticle-2720, Changes to Inspection Interval.

Specifically, for ISFSI sites meeting the conditions below, the canister inspection interval may be increased to maximum of 40 years without meeting the requirements in ASME CC N-860, Subsubparagraphs-2720(a)(1) and -2720(a)(3), which state that three consecutive inspections must be completed with consistent results. For this alternative, the ISFSI sites must meet both of the site conditions below.

1. the ISFSI site mush have a CISCC susceptibility ranking of 3 or below as determined using the criteria in EPRI-3002005371 (Ref. 9), and

2. all other requirements of ASME CC N-860 Subparagraph-2720(a) are satisfied.

Pre-Decisional for ACRS Consideration DG-3058 Page 11 Pre-Decisional for ACRS Consideration D. IMPLEMENTATION The NRC staff may use this regulatory guide as a reference in its regulatory processes, such as licensing, inspection, or enforcement. However, the NRC staff does not intend to use the guidance in this regulatory guide to support NRC staff actions in a manner that would constitute backfitting as that term is defined in 10 CFR 72.62, Backfitting, and as described in NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests (Ref. 296). The staff also does not intend to use the guidance to support NRC staff actions in a manner that constitutes forward fitting as that term is defined and described in Management Directive 8.4. The backfitting and forward fitting considerations in 10 CFR 72.62 and NRC Management Directive 8.4 apply to holders of general and specific licenses for ISFSIs and monitored retrievable storage installations issued under 10 CFR Part 72. However, the backfitting and forward fitting considerations in 10 CFR 72.62 and NRC Management Directive 8.4 do not apply to CoC holders. If a licensee believes that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfitting or forward fitting appeal with the NRC in accordance with the process in Management Directive 8.4.

Pre-Decisional for ACRS Consideration DG-3058 Page 12 Pre-Decisional for ACRS Consideration REFERENCES 1

1.

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, ASME, New York, N.Y.2

2.

U.S. Code of Federal Regulations (CFR), Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste, Part 72, Chapter I, Title 10, Energy

3.

U.S. Nuclear Regulatory Commission (NRC), NUREG-1927, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, Washington, DC.

4.

NRC, NUREG-2214, Managing Aging Processes In Storage (MAPS) Report, Washington, DC.

5.

NRC, Regulatory Guide 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III, Washington, DC.

6.

NRC Regulatory Guide 1.193, ASME Code Cases Not Approved for Use, Washington, DC.

7.

NRC Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Washington, DC.

8.

U.S. Code of Federal Regulations, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Energy.

8.9.

ASME Code Case N-860, Inspection Requirements and Evaluation Standards for Spent Nuclear Fuel Storage and Transportation Containment SystemsSection XI, Division 1;Section XI, Division 2, ASME B&PV Code Cases: Nuclear Components, Supplement 6, 2019 9.10.

Electric Power Research Institute (EPRI), Susceptibility Assessment Criteria for Chloride-Induced Stress Corrosion Cracking (CISCC) of Welded Stainless Steel Canisters for Dry Cask Storage Systems, EPRI-3002005371, Palo Alto, CA, 20153 1

Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

2 Copies of American Society of Mechanical Engineers (ASME) standards may be purchased from ASME, Two Park Avenue, New York, New York 10016-5990; telephone (800) 843-2763. Purchase information is available through the ASME Web-based store at http://www.asme.org/Codes/Publications/.

3 Copies of Electric Power Research Institute (EPRI) standards and reports may be purchased from EPRI, 3420 Hillview Ave., Palo Alto, CA 94304; telephone (800) 313-3774; fax (925) 609-1310.

Pre-Decisional for ACRS Consideration DG-3058 Page 13 Pre-Decisional for ACRS Consideration 10.11. NRC, Regulatory Guide 3.76, Implementation of Aging Management Requirements for Spent Fuel Storage Renewals, Washington, DC.

11.12. Nuclear Energy Institute (NEI), NEI 14-03, Revision 2, Format, Content and Implementation Guidance for Dry Cask Storage Operations-Based Aging Management, Washington, DC.,

December 2016, Agencywide Documents Access and Management System (ADAMS) Accession No. ML16356A2044 12.13. National Laboratories (SNL), SAND2020-13674, Analysis of Dust Samples Collected from an Inland ISFSI Site (Site A), December 20205 13.14. SNL, SAND2020-14144, Analysis of Dust Samples Collected from an Inland ISFSI Site (Site B), December 2020 14.15. SNL, SAND2022-10884, Analysis of Dust Samples Collected from a Near-Marine East Coast ISFSI Site (Site C), August 2022 15.16. NRC, Information Notice 2012-20, Potential for Chloride-Induced Stress Corrosion Cracking of Austenitic Stainless Steel and Maintenance of Dry Cask Storage Systems, issued November 14, 2012, Washington, DC.

16.17. J.W. Chi, W.J. Li, D.Z. Zhang, J.C. Zhang, Y.T. Lin, X.J. Shen, J.Y. Sun, J.M. Chen, X.Y.

Zhang, Y.M. Zhang, and W.X. Wang, Sea salt aerosols as a reactive surface for inorganic and organic acidic gases in the Arctic troposphere, Atmospheric Chemistry and Physics, Vol. 15, pp.

11341-11353, 2015 17.18. H.P. Leckie and H.H. Uhlig, Environmental factors affecting the critical potential for pitting in 18-8 stainless steel, Journal of the Electrochemical Society, Vol. 113, pp. 1262-1267, 1966 18.19. A.J.Sedriks, Corrosion of Stainless Steels, Second Edition, John Wiley, New York, NY, 1996 19.20. Z. Szklarska-Smialowska, Pitting and Crevice Corrosion, NACE International, Houston, TX, 2005 20.21. A.J.M.C. Cook, C. Padovani, A.J. Davenport, Effect of Nitrate and Sulphate of Atmospheric Corrosion of 304L and, 316L Stainless Steels, Journal of the Electrochemical Society, Vol.

164(4), pp. C148-C163, 2017 21.22. Rideout, S. and R. Mittelberg, Factors Controlling Chloride Stress Corrosion of Type 304 Stainless Steel, U.S. A.E.C. Report CONF-778-3, 1964 22.23. Uhlig, H.H. and E.W. Cook Jr. Mechanism of Inhibiting Stress Corrosion Cracking of 188 Stainless Steel in MgCl2 by Acetates and Nitrates, Journal of The Electrochemical Society, Volume 116(2) pp. 173-177, February 1969 4

Publications from the Nuclear Energy Institute (NEI) are available at its website: http://www.nei.org/ or by contacting the headquarters at Nuclear Energy Institute, 1776 I Street NW, Washington, DC 20006-3708, Phone: 202-739-8000, Fax 202-785-4019.

5 Sandia National Laboratories (SNL) reports are available electronically via the U.S. Department of Energy (DOE) Office of Scientific and Technical Information web site (www.osti.gov).

Pre-Decisional for ACRS Consideration DG-3058 Page 14 Pre-Decisional for ACRS Consideration 23.24. Bryant, R.E and J.B. Greer, Inhibition of chloride stress corrosion cracking of stainless alloys, Materials Protection and Performance, Volume 9, No. 11, pp.19-23, November, 1970

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NRC, Nuclear Regulatory Commission International Policy Statement, Federal Register, Vol. 79, No. 132, July 10, 2014, pp. 39415-39418, Washington, DC.

26.

NRC, Management Directive 6.6, Regulatory Guides, Washington, DC.

24.27. IAEA Safety Standards, Specific Safety Requirements No. SSR-4, Safety of Nuclear Fuel Cycle Facilities, International Atomic Energy Agency, Vienna, Austria6

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IAEA Safety Standards, Specific Safety Guide SSG-15, Storage of Spent Nuclear Fuel, International Atomic Energy Agency, Vienna, Austria.

25.29. NRC, Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, Washington, DC

26.

NRC, Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests 6

Copies of International Atomic Energy Agency (IAEA) documents may be obtained through their Web site:

WWW.IAEA.Org/ or by writing the International Atomic Energy Agency, P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria.

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