ML23269A225

Rev 0 Guidance on Ticap for Non-LWRs Rlso

ML23269A225
Person / Time
Issue date:	03/22/2024
From:	Anders Gilbertson NRC/NRR/DANU/UARP
To:
References
DG-1404 RG 1.253 Rev 0
Download: ML23269A225 (51)

Text

U.S. NUCLEAR REGULATORY COMMISSION REGULATORY GUIDE 1.253, Revision 0

Issue Date: March 2024 Technical Lead: Anders Gilbertson

GUIDANCE FOR A TECHNOLOGY-INCLUSIVE CONTENT-OF-APPLICATION METHODOLOGY TO INFORM THE LICENSING BASIS AND CONTENT OF APPLICATIONS FOR LICENSES, CERTIFICATIONS, AND APPROVALS FOR NON-LIGHT-WATER REACTORS

A. INTRODUCTION

Purpose

This regulatory guide (RG) describes an approach that is accept able to the staff of the U.S. Nuclear Regulatory Commission (NRC) for using a technology -inclusive content-of-application methodology to inform specific portions of the safety analysis report (SAR) included as part of a non-light-water reactor (non-LWR) license application. Specifically, this RG endorses the methodology described in Nuclear Energy Institute (NEI) report NEI 21-07, Revision 1, Technology Inclusive Guidance for Non-Light Water Reactors, Safety Analysis Report Content: For Applicants Using the NEI 18-04 Methodology issued February 2022 (Ref. 1), with clarific ations and additions, where applicable, as one acceptable process for use in developing certain portion s of the SAR for an application for a non-LWR construction permit (CP) or operating license (OL) under Ti tle 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utiliz ation Facilities (Ref. 2), or for a combined license (COL), or design certification (DC) unde r 10 CFR Part 52, Licenses, Commented [A1]: NRC-2022-0074-DRAFT-0006-2 through Certifications, and Approvals for Nuclear Power Plants (Ref. 3 ). As of the date of this RG, the NRC is 0006-9 developing a rule to amend Parts 50 and 52 (RIN 3150-Al66). The NRC staff notes this RG may need to be updated to conform to changes to Parts 50 and 52, if any, ad opted through that rulemaking. Further, as of the date of this RG, the NRC is developing an optional perfo rmance-based, technology-inclusive regulatory framework for licensing nuclear power plants designa ted as 10 CFR Part 53, Licensing and Regulation of Advanced Nuclear Reactors, (RIN 3150-AK31) and a nticipates that this RG will be updated after promulgation of those regulations to address cont ent of application considerations specific to the licensing processes in this framework.

Applicability

This RG applies to designers of non-LWRs and applicants for per mits, licenses, and certifications under 10 CFR Part 50 and 10 CFR Part 52 for such reactors. Upon conclusion of the rulemaking Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site in the NRC Library at https://www.nrc.gov/reading-rm/doc-collections/reg-guides/index.html, under Document Collections, in Regulatory Guides, at https://www.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html. During the development process of new guides suggestions shou ld be submitted within the comment period for immediate consideration. Suggestions received outside of the comment period will be considered if practical to do so or may be considered for future updates.

Electronic copies of this RG, previous versions of RGs, and other recently issued guides are also available through the NRCs public Web site in the NRC Library at https://www.nrc.gov/reading-rm/doc-collections/reg-guides/index.html, under Document Collectio ns, in Regulatory Guides.

This RG is also available throu gh the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession Number (No.) ML23269A222. The regulator y analysis may be found in ADAMS under Accession No. ML22076A002. The associated draft guide DG-1404, Revision 0 and Revision 1, may be found in ADAMS un der Accession No. ML22076A003, and ML23194A194, respectively. The staff responses to the public comments on DG-1404 may be found under ADAMS Accession No. ML23269A223.

underway to amend 10 CFR Parts 50 and 52, the NRC may update th is RG, if necessary, to conform to changes to Parts 50 and 52 adopted through that rulemaking. The NRC staff envisions that the approach in this RG will also support the technology-inclusive, risk-inf ormed, and performance-based licensing framework that is now under development and currently designate d as 10 CFR Part 53 (RIN 3150-AK31). The NRC staff plans to update this RG to reflect these r egulations after a final rule is promulgated to reflect any additional guidance unique to the content of app lications under those regulations.

Applicable Regulations1

• 10 CFR Part 50 contains regulations for licensing production an d utilization facilities.

o 10 CFR 50.34, Contents of applications; technical information, describes the minimum information required for (1) preliminary safety analysis reports supporting CP applications and (2) final safety analysis reports (FSARs) supporting OL applica tions.

• 10 CFR Part 52 governs the issuance of DCs, and COLs for nuclea r power facilities. Commented [A2]: NRC-2022-0074-DRAFT-0006-2 through 0006-9 o 10 CFR 52.47, Contents of applications; technical information, describes the information required to be included in FSARs supporting applications for st andard DCs.

o 10 CFR 52.79, Contents of applications; technical information in final safety analysis report, describes the information required to be included in FSARs supp orting applications for COLs.

Related Guidance

• Policy Statement on the Regulation of Advanced Reactors (Volu me 73 of the Federal Register (FR), page 60612 (73 FR 60612); October 14, 2008) (Ref. 4), est ablishes the Commissions policy for advanced reactor designs to protect the environment and public health and safety and promote the common defense and security.

• RG 1.70, Standard Format and Content of Safety Analysis Report s for Nuclear Power Plants: LWR Edition (Ref. 5), provides detailed guidance to the writers of SARs to allow for the standardization of information the NRC needs for reviewing CPs and OL applicati ons.

• RG 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e) (Ref. 6), provides methods the NRC staff finds acceptable for complying with the provisions of 10 CFR 50.71(e), which requires periodic developm ent of updates to FSARs.

• RG 1.206, Applications for Nuclear Power Plants (Ref. 7), pro vides guidance on the format and content of applications for licenses, certifications, and appro vals for nuclear power plants submitted to the NRC under 10 CFR Part 52. RG 1.206 specifies the informa tion to be included in an application for a light-water reactor (LWR), although the guida nce may also be generally useful for non-LWR applications.

1 The staff notes that for advanced reactors, the NRC will det ermine the applicability of specific technical requirements in the regulations, or the need to define additional technical requirements based on the safety assessments for a particular design, on a case-by-case basis. Applicants seeking to use the NEI 18-04 approach in the development of an application for a standard design approval (SDA) or manufacturing license (ML) should engage in pre-application dialogue with the NRC.

RG 1.253, Page 2

• RG 1.232, Guidance for Developing Principal Design Criteria fo r Non-Light-Water Reactors (Ref. 8), contains guidance on adapting the general design crit eria in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, to non-LWR designs. Non-LWR designers, applicants, and licensees may use this guidance to develop prin cipal design criteria (PDC) for any non-LWR designs, as required by the applicable NRC regulations. RG 1.232 also contains guidance for modifying and supplementing the general design criteria to develop PDC for two types of non-LWR technologies: sodium-cooled fast reactors and modular h igh-temperature gas-cooled reactors.

• RG 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applic ations for Licenses, Certifications, and Approvals for Non-Light-Water Reactors (Ref. 9), contains the NRCs endorsement of the Licensing Modernization Project (LMP) methodology in NEI 18-04, Revision 1, Risk-Informed Performance-Based Guidance for Non-Light Water Reactor Licensin g Basis Development, (Ref. 10) for selecting licensing-basis events (LBEs); classifying structures, systems, and components (SSCs);

and assessing the adequacy of defense in depth (DID). Non-LWR r eactor designers, applicants, and licensees may use this guidance to develop the content of their applications for non-LWR designs.

Specifically, for applicants following the guidance in NEI 21-0 7, Revision 1, the LMP methodology is the baseline approach for developing the application.

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 techniqu es that the staff uses in evaluating specific issues or postulated events, and to describe information that t he staff needs in its review of applications for permits and licenses. Regulatory guides are not NRC regulat ions and compliance with them is not required. Methods and solutions that differ from those set fort h 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 manda tory information collections in 10 CFR Parts 50 and 52 that are subject to the Paperwork Reduct ion Act of 1995 (44U.S.C.3501 et.

seq.). These information collections were approved by the Offic e of Management and Budget (OMB),

under control numbers 3150-0011 and 3150-0151, respectively. Se nd comments regarding this information collection to the FOIA, Library, and Information Co llections 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 Offic er of the Nuclear Regulatory Commission, 725 17th Street, NW, Washington,DC, 20503.

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not requir ed to respond to, a collection of information unless the document requesting or requiring the col lection displays a currently valid OMB control number.

RG 1.253, Page 3 TABLE OF CONTENTS

Contents A. INTRODUCTION................................................................................................................................................................................................................ 1 B. DISCUSSION.......................................................................................................................................................................................................................... 5 C. STAFF REGULATORY GUIDANCE..................................................................................................................................................................... 10 D. IMPLEMENTATION...................................................................................................................................................................................................... 24 ACRONYMS/ABBREVIATIONS..................................................................................................................................................................................... 25 REFERENCES........................................................................................................................................................................................................................... 26 Appendix A Acceptability of a Probabilistic Risk Assessment Tha t Supports a Non-Light-Water Reactor Construction Permit Application Based on the Licensing Modernization Project Methodology A-1

RG 1.253, Page 4 B. DISCUSSION

Reason for Issuance

This RG provides the NRC staffs guidance on using a technology -inclusive content-of-application methodology to develop specific portions of the SAR included as part of a non-LWR license application. Specifically, this RG endorses the methodology described in NEI 21-07, Revision 1,2 as one acceptable method for use in developing certain portion s of the SAR for an application for a non-LWR CP or OL under 10 CFR Part 50, or COL or DC under 10 CFR Part 52, with Commented [A3]: NRC-2022-0074-DRAFT-0006-2 through clarifications and additions described below. The technology-in clusive methodology in NEI 21-07, 0006-9 Revision 1, provides a common approach for the development of those portions of the SAR that reflect the outcomes and insights from the implementation of the Licens ing Modernization Project (LMP) methodology as described in NEI 18-04, Revision 1, and endorsed by the NRC in Regulatory Guide 1.233. The applicant is also responsible for demonstrating comp liance with all applicable regulations and Commented [A4]: NRC-2022-0074-DRAFT-0006-10 may request exemptions, as appropriate, to establish the licens ing basis for the design.3

=

Background===

As the NRC prepares to review and regulate a new generation of non-LWRs, the staff has recognized both the need to establish a flexible regulatory fra mework and the benefits of such a framework. The NRC described its efforts to prepare for possible licensing of non-LWR technologies in NRC Vision and Strategy: Safely Achieving Effective and Effici ent Non-Light Water Reactor Mission Readiness, issued December 2016 (Ref. 11). In NRC Non-Light W ater Reactor Near-Term Implementation Action Plans, issued July 2017 (Ref. 12), and NRC Non-Light Water Reactor Mid-Term and Long-Term Implementation Action Plans, issued Jul y 2017 (Ref. 13), the NRC staff identified specific activities the agency would conduct in the near-term, mid-term, and long-term timeframes. In addition, the Commission encouraged the use of a performance-based, technology-inclusive licensing framework for small modular reac tors in SRM-COMGBJ-10-0004/COMGEA-10-0001, Staff Requirements COMGBJ-10-0004/COMGEA-10-0001Use of Risk Insights to Enhance S afety Focus of Small Modular Reactor Reviews, dated August 31, 2010 (Ref. 14), and SRM-SECY-11-0024, Staff RequirementsSECY-11-0024Use of Risk Insights to Enhance the S afety Focus of Small Modular Reactor Reviews, dated May 11, 2011 (Ref. 15). The NRC staff b elieves that this approach is appropriate to apply to the guidance development for non-LWRs. 16

Efforts to establish a risk-informed, performance-based, techno logy-inclusive regulatory framework for non-LWRs included the development of several key guidance documents. These include guidance on adapting the general design criteria in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, for developing principal de sign criteria for non-LWR designs documented in RG 1.232. In addition, these also include the LMP described in the industry-developed guidance NEI 18-04, Revision 1, issued August 2019. NEI 18-04, Revision 1, was endorsed by the NRC in RG 1.233. NEI 18-04, Revision 1, provides a systematic, risk -informed, and technology-inclusive process for developing key inputs for the content of applicatio ns to improve the understanding of the

2 The NRC encourages an LWR applicant proposing to use the risk-informed, performance-based process described in NEI 21-07, Revision 1, to engage in pre-application discussions with the NRC to provide information to the staff on its intended implementation of the NEI 21-07, Revision 1 methodology for its design.

3 The staff has provided guidance on which regulations apply to non-LWRs in Appendix B of the advanced reactor content of application project (ARCAP) Roadmap interim staff guidance (ISG) document, DANU-ISG-2022-01, Review of Risk Informed, Technology Inclusive Advanced Reactor ApplicationsRo admap (Ref. 16).

RG 1.253, Page 5 safety and risk significance of system designs and the relation ship of system functions to overall facility safety, specifically for non-LWR designs.

A key element of this new and flexible regulatory framework is to standardize the development of non-LWR application content to promote uniformity among applica nts, support staff review consistency and predictability, and provide a well-defined base for evaluat ing proposed changes in review scope and requirements. The development of an application for an NRC lice nse, permit, certification, or approval is a major undertaking, in that an applicant must provide sufficie nt information to support the agencys safety findings. The information and level of detail needed wil l vary according to whether an application is for a CP, DC, OL, COL, or other action. Commented [A5]: NRC-2022-0074-DRAFT-0006-2 through 0006-9 The NRC staff has had success with a standard content-of-applic ation methodology for large LWRs. RG 1.70, issued in the 1970s, and RG 1.206, issued in 200 7 and revised in 2018, reflect the NRCs efforts to standardize the format and content of LWR appl ications. Guidance documents such as these and numerous others on specific technical areas address t he suggested scope and level of detail for those applications.4 17

To standardize the development of advanced reactor application content, the staff has focused on two projects:

• advanced reactor content of application project (ARCAP)

• technology-inclusive content of application project (TICAP)

ARCAP is an NRC-led activity intended to provide guidance for a complete non-LWR application under either 10 CFR Part 50 or 10 CFR Part 52, and eventually the technology-inclusive, performance-based licensing framework for which a rule is now b eing developed as 10 CFR Part 53. As a result, ARCAP is broad, encompassing several industry-led and N RC-led guidance development efforts that aim to promote consistency in developing applications. As described in the ARCAP Roadmap ISG, a complete non-LWR application should include, among other things, an SAR that includes proposed Commented [A6]: NRC-2022-0074-DRAFT-0006-11 technical specifications, an emergency plan, and other informat ion such as physical security plans. Commented [A7]: NRC-2022-0074-DRAFT-0006-12

TICAP is an industry led guidance activity focused on the scope and depth of information to include in the portions of an SAR that address the implementati on of the LMP methodology described in NEI 18-04, Revision 1, and endorsed by the NRC in Regulatory Gu ide 1.233. By focusing on those Commented [A8]: NRC-2022-0074-DRAFT-0006-10 aspects of the facility design most relevant to the risks posed by non-LWR technologies, including design features and human actions, the TICAP guidance will help applic ants provide sufficient information on the design and programmatic controls, while obviating the need for excessive detail in less important areas. The specific portions of the SAR within the scope of TIC AP are described below in more detail.

The ARCAP guidance encompasses and supplements the TICAP guidan ce. In particular, the ARCAP documents address areas of the SAR that are outside the scope o f the TICAP guidance (i.e., not covered by the LMP process), such as technical specifications, control of routine plant effluents, control of occupational exposure, etc.

4 In addition, NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:

LWR Edition (Ref. 17), provides guidance to the staff on how to review applications.

RG 1.253, Page 6 As a result of extensive public discussion on TICAP and ARCAP w ith industry and other external stakeholders, the NRC has proposed a 12-chapter structure for t he SAR for a non-LWR application. In contrast, the SAR for large LWRs, as described in RG 1.206, has 19 chapters. The staff on occasion adds guidance to its structure to discuss matters not evaluated in o ther review guidance chapters. The 12 chapters proposed for an advanced reactor SAR, consistent wi th ARCAP/TICAP guidance and the methodology in this RG, are as follows:

Chapter 1, General Plant and Site Description, and Overview of the Safety Analysis Chapter 2, Methodologies, Analyses, and Site Evaluations 5 Chapter 3, Licensing-Basis Events Chapter 4, Integrated Evaluations Chapter 5, Safety Functions, Design Criteria, and Structure, S ystem, and Component Safety Classifications Chapter 6, Safety-Related (SR) Structure, System, and Componen t Criteria and Capabilities Chapter 7, Non-Safety-Related with Special Treatment (NSRST) S tructure, System, and Component Criteria and Capabilities Chapter 8, Plant Programs Chapter 9, Control of Routine Plant Radioactive Effluents, Pla nt Contamination, and Solid Waste Chapter 10, Control of Occupational Dose Chapter 11, Organization and Human-System Considerations Chapter 12, Post-Construction Inspections, Testing, and Analys is Programs Other documents incorporated by reference into the SAR (e.g., e mergency plan)

For applications that follow the SAR structure above, the scop e of TICAP as described in NEI 21-07, Revision 1, includes the first eight of these chapte rs (i.e., those informed by the LMP process).6 Figure 1 illustrates the nexus between ARCAP, TICAP, and other guidance for an advanced reactor application.7 Commented [A9]: NRC-2022-0074-DRAFT-0006-12

5 The LMP process does not specifically address site evaluatio ns, therefore, NEI 21-07, Revision 1, does not address this aspect for an advanced reactor application. However, PRA may be used to inform the design considerations of external hazards associated with potential sites. Therefore, the discuss ion on SAR content and organization provided in this TICAP RG is from the perspective of the overall application and includes guidance for aspects considered outside of the LMP process. ARCAP Chapter 2, also includes such guidance on application content on site evaluations that should be included in Chapter 2 of the SAR.

6 The NRC highly encourages pre-application engagement from applicants that plan to use the methodology in NEI-21-07, Revision 1, but rely on a different SAR structure than the 12-c hapter approach described in this RG and addressed in ARCAP. Similarly, applicants following the 12-chapter SAR struc ture but not using the LMP approach of NEI 21-07, Revision 1, should engage the NRC staff early to ensure the application contains all the information required by regulations and to optimize application reviews. The Commissions 2008 Policy Statement on the Regulation of Advanced Reactors, highlights the importance of pre-application activities.

7 Requirements for the contents of a final safety analysis report (FSAR) are provided in 10 CFR 50.34(b) and include items such as proposed technical specifications and emergency plans, as well as other technical and programmatic contents listed therein. It should be noted that items such as technical specif ications and emergency plans may be incorporated by reference in the FSAR but are controlled by change processes other than 10 CFR 50.59 for OLs. For example, changes to the technical specifications, which are part of the license, require a licens e amendment, and emergency plan changes are controlled by 10 CFR 50.54(q).

RG 1.253, Page 7 Figure 1. Relationship between ARCAP, TICAP, and the content of an application

Documents Endorsed in this Guide

After completing the TICAP efforts, NEI documented the results of the project as guidance in NEI 21-07, Revision 1 and submitted that guidance to the NRC fo r review and endorsement. The purpose of this RG is twofold:

1. To endorse NEI 21-07, Revision 1, with clarifications and addit ions. NEI 21-07, Revision 1, describes one acceptable approach for determining the scope and level of detail for the development of structured application content associated with t he first eight chapters of the SAR. NEI 21-07, Revision 1, follows the LMP guidance and system atically describes the selection of LBEs; the classification and special treatment of SSCs; and the assessment of DID adequacy. Where applicable, this RG describes additional po ints of emphasis or further details relevant to the SAR sect ions discussed in NEI 21-07, Re vision 1, and endorsed by this RG.

2. To provide additional guidance and information outside the scop e of the LMP methodology and NEI 21-07, Revision 1, that the NRC staff has determined is also relevant and should be included as part of the application content related to the firs t eight chapters of the SAR.

RG 1.253, Page 8 Accordingly, this RG endorses NEI 21-07, Revision 1, with clari fications and additions, as one acceptable approach for use in developing certain portions of a n SAR for a license, permit, or certification application to the NRC for a non-LWR using the methodology endo rsed in RG 1.233. Additional details for each chapter appear in the corresponding section below.

In summary, the guidance in NEI 21-07, Revision 1, focuses on t he portions of the SAR containing material addressed using the LMP process in NEI 18-0 4, Revision 1. The guidance in NEI 21-07, Revision 1, with the clarifications and additions de scribed in this regulatory guide, will promote the submission of complete information to the NRC and e nsure that application content is commensurate with the risk significance and complexity of the d esign and associated safety analysis.

NEI 21-07, Revision 1, provides a standardized content developm ent process to facilitate efficient SAR preparation by the applicant, NRC review of the application, an d, if approved by the NRC, maintenance by the licensee. The guidance in NEI 21-07, Revision 1, should optimize the scope, content, and level of detail of each application, based on the risk significance and complexity of the design and associated safety analysis and the nexus between design elements and publi c health and safety.

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 technolog ies. The IAEA develops Safety Requirements and Safety Guides for protecting people and the en vironment 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 Requ irements and Safety Guides pursuant to the Commissions International Policy Statement (Ref. 18) an d Management Directive and Handbook 6.6, Regulatory Guides (Ref. 19).

The following IAEA Safety Requirements were considered in the d evelopment of this Regulatory Guide:

• IAEA Specific Safety Requirements No. SSR-2/1, Safety of Nuclear Power Plants: Design, issued 2016 (Ref. 20)

Documents Discussed in Staff Regulatory Guidance

This RG endorses, in part, the use of one or more third-party g uidance documents. These third-party guidance documents may contain references to other codes, standards, or third-party guidance documents (secondary references). If a secondary reference ha s itself been incorporated by reference into NRC regulations as a requirement, then licensees and appli cants must comply with that standard as set forth in the regulation. If the secondary reference has bee n endorsed in a RG as an acceptable approach for meeting an NRC requirement, then the standard cons titutes a method acceptable to the NRC staff for meeting that regulatory requirement as described in t he 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 justif ied, consistent with current regulatory practice, and consistent with applicable NRC requirements.

RG 1.253, Page 9 C. STAFF REGULATORY GUIDANCE

This RG endorses the methodology described in NEI 21-07, Revis ion 1, as one acceptable method for use in developing certain portions of the SAR for an application for a non-LWR CP or OL under 10 CFR Part 50, or a COL or DC under 10 CFR Part 52. Howe ver, the NRC staff provides Commented [A10]: NRC-2022-0074-DRAFT-0006-2 clarifications and additions to certain statements in NEI 21-07, Revision 1, as discussed below. through 0006-9

The guidance in this RG on the SAR scope, content, and level of detail is based on the appropriate level of design-specific information that should be provided in an application to the NRC to demonstrate that the facility design meets the regulatory stand ards for adequate protection of public health and safety. To provide effective and efficient technology-inclu sive content guidance while ensuring the current application content requirements are met, this guidance describes an LMP-based safety analysis.

The NRC highly encourages pre-application engagement between ap plicants and the staff to promote common understanding of proposed regulatory approaches, unique and novel designs, and technical issues, and to optimize resources and review schedules, especia lly for non-LMP-based applications.

The following sections describe the NRCs endorsement (with cl arifications and additions, where applicable) of the corresponding chapters in NEI 21-07, Revisio n 1. In general, NEI 21-07, Revision 1, recommends that applicants first present the overall safety ana lysis for the reactor and then give supporting design and operational details in subsequent chapter s. The staff notes that the methods, approaches, and data described in the regulatory guidance posit ions below are considered guidance and not requirements. However, in addition to presenting the overal l safety analysis specific to their designs, applicants are required by the content of application requireme nts in Parts 50 and 52 to present a complete licensing basis by demonstrating compliance with applicable reg ulations, including any exemptions, where necessary, along with sufficient justification for each e xemption. The suitability of such an exemption would be design-dependent, its justification would be the responsibility of the applicant, and the NRC would evaluate it on a case-specific basis.

NEI 21-07 explicitly addresses several licensing pathways: a co mbined license (COL) under 10 CFR Part 52 Subpart C; a design certification (DC) under 10 CFR Part 52 Subpart B; and a two-step license (CP/OL) under 10 CFR Part 50. An applicant using a lice nsing pathway other than one explicitly covered in NEI 21-07 may base the SAR content on the licensing pathway covered by NEI 21-07 and most similar to the approach it is using. For example, an appli cant seeking a manufacturing license (ML) under 10 CFR Part 52 Subpart F or standard design approval (SDA ) under 10 CFR Part 52 Subpart E may start with the guidance for a DC and make the necessary modific ations to address its specific proposal and the applicable regulations. While such an approach reflects the similarity in the required contents of applications for the various licensing pathways in Parts 50 and 52, it will be up to the applicant to justify the guidance as applied and adjusted to address the scope of th e application and to address the differences in the regulation for an ML or SDA application using the LMP me thodology. As noted in Footnote 1, an ML or SDA applicant seeking to use RG 1.253 guidance should eng age in a pre-application dialogue with the NRC. Commented [A11]: NRC-2022-0074-DRAFT-0006-2 through 0006-9

1. Introduction and Development of Guidance

Section A, Introduction, and Section B, Development of Guid ance, of NEI 21-07, Revision 1, discuss the documents purpose, background, scope, and organiza tion, as well as the development of the guidance, an outline of the SAR, general instructions for use o f the guidance, alternate licensing paths, two-step licensing (CP/OL), and DCs. Section C, SAR Content Gu idance, gives specific guidance on developing SAR content for a COL, with supplemental information for CP/OL and DC applications, where these differ from COL applications.

RG 1.253, Page 10 NEI 21-07, Revision 1, Section B.3, General Instructions for U se of the Guidance, states the following:

Italicized text provides background information for context and perspective. It is intended to provide readily accessible supporting information, but the i talicized text does not require direct action on the part of the applicant. Information that is general in nature (e.g., general goals for level of detail, expectations for orga nization) will also be provided in italic.

C.1 Staff Position: NEI 21-07 Sections A and B provide acceptable background asso ciated with TICAP guidance development with the following clarification:

a. The staff considers all discussion in NEI 21-07, Revision 1, to constitute guidance and not requirements; therefore, the staff considers the italicized tex t in NEI 21-07, Revision 1, to be part of the guidance and not simply background and context.

2. General Plant and Site Description and Overview of the Safety Analysis

Section C.1 of NEI 21-07, Revision 1, provides guidance for dev eloping baseline information related to the plant description, the site description, the saf ety analysis based on the LMP methodology, and a summary of reference or source materials.

As described in NEI 21-07, Revision 1, the information in Chap ter 1 of an SAR that follows NEI 21-07, Revision 1, should give the reviewer a basic underst anding of the overall facility, such as the type of permit, license, certification, or approval requested; the number of reactor units; a brief description of the proposed plant location; and the type of adv anced reactor being proposed. The site description should provide an overview of the actual physical, environmental, and demographic features of the proposed site, and how they relate to the safety analysi s. For example, the site description should include geological, demographic, seismological, hydrological, a nd meteorological characteristics of the site and its vicinity.

In NEI 21-07, Revision 1, NEI defines the affirmative safety case as a collection of technical and programmatic information that demonstrates that the design meets the performance objectives of the technology-inclusive fundamental safety functions during design -specific anticipated operational occurrences (AOOs), design-basis events (DBEs), beyond-design-b asis events (BDBEs), and design-basis accidents (DBAs). As described in NEI 21-07, Revision 1, sectio n A.3., the affirmative safety case should do the following:

• Identify design-specific safety functions that are adequately p erformed by design-specific SSCs.

• Establish design-specific features to provide reasonable assura nce that credited SSC functions are reliably performed and to demonstrate DID adequacy.

C.2 Staff Position: NEI 21-07, Revision 1, Section C.1, provides an acceptable me thod for developing baseline information related to the plant descriptio n, the site description, the overall safety analysis based on the LMP methodology, and a summary of referen ce or source materials with clarifications and additions as noted below.

a. Clarification: NEI 21-07, Revision 1, includes use of the terms affirmative safety case, safety case, and licensing case. To avoid confusion and potential u nforeseen consequences, applicants using NEI 21-07, Revision 1, should instead continue to use the established terminology in the

RG 1.253, Page 11 current regulatory framework, including use of safety analysis and licensing basis.8

b. Clarification: The LMP methodology endorsed in RG 1.233 by its nature addresses off-normal Commented [A12]: NRC-2022-0074-DRAFT-0006-13 conditions rather than normal operation. Applicants using NEI 2 1-07, Revision 1 to develop their SARs should also include additional information in parts of the SAR not derived from the LMP to describe and analyze normal operation. In addition, an applican t using NEI 21-07, Revision 1, is also responsible for demonstrating compliance with all applicable re gulations, including exemptions, as necessary, with sufficient justification. Appendix B to the ARC AP Roadmap ISG contains staff guidance on which regulations apply to non-LWRs.

c. Addition: In addition to the information identified in NEI 21-0 7, Revision 1, Section C.1.1.2, on intended use of the reactor, applicants should also provide the nature (e.g., physical form) and inventory of contained radioactive materials in Chapter 1 or other appropriate sections of the SAR. Commented [A13]: NRC-2022-0074-DRAFT-0006-15

d. Addition: NEI 21-07 calls for the application to include discus sions of the analyses of the potential radiological consequences from various event sequences identifi ed from the PRA and related assessments. NEI 18-04 includes a specific question to be addre ssed during the integrated decision-making process related to the assessment of cliff edge effects. ASME/ANS RA-S-1.4-2021 addresses possible cliff-edge effects in areas such as seismic and flooding hazards. Any design requirements or special treatment of SSCs to prevent or mitigat e cliff-edge effects should be included in the SSC specific descriptions in subsequent SAR chapters.

3. Methodologies, Analyses, and Site Evaluations

Section C.2 of NEI 21-07, Revision 1, presents guidance on the information to be included in the SAR on certain analyses and analytical tools (methodologies) us ed to identify LBEs, evaluate their consequences, or assess the performance of SSCs that are either safety-related (SR) or non-safety-related with special treatment (NSRST). The amount of information direc tly stated in this chapter, as opposed to incorporated by reference, could depend upon the extent of pre-licensing interactions between the applicant and the NRC, particularly interactions resulting in s taff reviews and approvals (i.e., topical reports) and the extent to which the application relies on anot her license or certification (e.g., a COL referencing a certified design).

The information to be provided in Chapter 2 of an SAR followin g NEI 21-07, Revision 1, is primarily cross-cutting information or evaluations that support multiple LBEs or SSCs and provide a foundation for more specific information and analysis results g iven in other chapters of the SAR. The information provided in Chapter 2 should focus on the probabili stic risk assessment (PRA), source term analysis, DBA analytical methods, and other methodologies and a nalyses (e.g., civil and structural analysis, piping analysis, electrical load analysis, stress ana lysis, criticality analysis, thermal-hydraulic analysis, environmental qualification analysis, and dispersion modeling) that are pertinent to the LMP-based safety analysis.

When complete and final design information is not available at the CP application stage, the plant design and the associated PRA are considered preliminary, since they are less mature than they are at the OL stage. Therefore, the description of the PRA in a CP applica tion should be a high-level overview or summary of topics such as the quality, scope, uses, and accepta bility of the PRA. The applicant should provide justification that the PRA has been performed in such a way that the PRA results are reasonable

8 The NRC staff notes that neither the LMP methodology in NEI 18-04 nor the staff endorsement of that methodology in RG 1.233 use these terms, and no NRC regulation or guidance defines them. These terms are unnecessary to implement the LMP approach.

RG 1.253, Page 12 given the level of maturity of the design, and that the SAR pro vides sufficient information to support the CP findings. The applicant should also include any necessary co mmitments to upgrade and maintain the PRA so that its completion status at the OL stage is consistent with its intended uses. For a 10 CFR Part 52 application, the level of detail of the PRA in the application should be sufficient to meet the requirements in 10 CFR Part 52 that the SAR include a descr iption of the PRA and its results.

The PRA is a model that provides an integrated assessment of t he risk to the public from the nuclear power plant. The PRA identifies and assesses the source s of radionuclides in the plant and the various plant operating states which, for example, include full power, low power, and shutdown conditions for reactors. Chapter 2 of an SAR following NEI 21-07, Revision 1, describes the PRA at a summary level, addressing its scope, methodology, and pedigree (e.g., technical acceptability, peer review). RG 1.247 (for trial use), Acceptability of Probabilis tic Risk Assessment Results for Advanced Non-Light Water Reactor Risk-Informed Activities (Ref. 21), en dorses with exceptions and clarifications the American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) non-LWR PRA standard, ASME/ANS RA-S-1.4-2021, Probabilis tic Risk Assessment Standard for Advanced Non-Light Water Reactor Nuclear Power Plants, (Re f. 22) and endorses with no exceptions and clarifications NEI 20-09, Performance of PRA Pe er Reviews Using the ASME/ANS Advanced Non-LWR PRA Standard (Ref. 23).

RG 1.247 describes one approach acceptable to the NRC staff for determining whether a PRA used to support an application is sufficient to provide confide nce in the results, such that the PRA can be used in regulatory decision making for non-LWRs. References may be included to other SAR chapters Commented [A14]: NRC-2022-0074-DRAFT-0006-17 that discuss the use of the PRA and its results (e.g., selectio n of LBEs, evaluation of LBE risk significance against the LMP frequency-consequence targets (Fig ure 3-1 of NEI 18-04, Revision 1),

determination of integrated risk and comparison to cumulative r isk metrics, PRA uncertainties and assessment of DID adequacy, PRA safety functions, SSC safety classification, and reliability and capability targets).9

In Chapter 2 of an SAR following NEI 21-07, Revision 1, the ap plicant provides information on event sequence source terms specific to its design that is used in the LBE consequence analyses. The source term information should cover all radioactive material i nventories and include the type, quantity, and timing of the release of radioactive material from the faci lity during LBEs. This chapter should include analysis methodologies, assumptions, bases, and justifi cations associated with transport of radioactive material from its point of origin to the accessible environment. For an LMP-based safety analysis, the application should include the use of a mechanist ic source term, consistent with the advanced non-LWR PRA standard definition (see Appendix A to NEI 21-07, Revision 1). Mechanistic source term information that is common to some or all of the ev ents considered for the plant may be given in Chapter 2 of an SAR following NEI 21-07, Revision 1, rather than repeated for each event. This information may include references to fuel qualification and pe rformance topical reports and the associated NRC safety evaluations, as discussed in the suppleme ntary information under Chapter 5.

C.3 Staff Position: Section C.2 of NEI 21-07, Revision 1, describes an acceptabl e method for developing baseline information related to the PRA (i.e., an ov erview of the PRA), source term analysis, DBA analytical methods, and other methodologies and analyses pe rtinent to the LMP-based safety analysis. Section C.2 of NEI 21-07, Revision 1 provides accepta ble guidance on the discussion of the software and analytical tools used to perform the event sequenc e modeling and quantification, determine the mechanistic source terms, and perform radiological conseque nce evaluations for the LBEs and DBAs

9 The NRC encourages an LWR applicant that proposes to use NEI 21-07, Revision 1 to engage with the staff in pre-application discussions on its use of PRA tools and techniques during implementation of the LMP process and the development of SAR content for its design.

RG 1.253, Page 13 listed in Section C.3 of NEI 21-07, Revision 1 and the cumulati ve dose and risk calculations in Section C.4.1 of NEI 21-07, Revision 1. Section C.2 of NEI 21-0 7, Revision 1 also specifies that the applicant should identify the methods used, describe at a high level how they are applied to the radiological consequence evaluations, and describe the site cha racteristics modeled or site parameters postulated in the radiological consequence evaluations. The fol lowing clarifications and additions are included:

Commented [A15]: NRC-2022-0074-DRAFT-0006-18

a. Clarification: NEI 21-07, Revision 1 calls for a discussion of how the NRC RG that endorses the non-LWR PRA standard was implemented (pending finalization of t he RG). This regulatory guide has subsequently been issued as RG 1.247 for trial use.

b. Addition: In addition to the information that NEI 21-07, Revisi on 1, states applicants should include in SAR Chapter 2, the SAR Chapter 2 should discuss the analysis methods and assumptions for the total calculated radiological consequence dose at the EAB, the outer boundary of the low-population zone (LPZ), and the control room (if required, e.g., if operato r actions are relied upon for safety-significant functions) to demonstrate that the facility meets the requirements of 10 CFR 50.34(a)(1)(ii)(D) or 10 CFR 52.79(a)(1)(vi) and the PDC for the control room (if applicable). Although an applicant is free to propose different approaches, two possible options for addressing these assessments based on the outcome of the LMP ap proach include: Commented [A16]: NRC-2022-0074-DRAFT-0006-20

(1) Option 1: Use the DBA dose consequence results from an LMP-base d approach to establish the acceptability of the EAB and LPZ. As described in RG 1.233, the DBA analysis under an LMP-based approach is a deterministic, conservative analysis th at is analogous to the DBA analyses performed for new LWRs and operating reactors. Under this option, depending on the nature of the DBA, the application may need to include an exemp tion from the regulations in 10 CFR 50.34 or 10 CFR 52.79 that require an assumed major acc ident10 to demonstrate containment performance and to confirm that the EAB and LPZ dos es are below the reference values in the regulations. An applicant is responsible for just ifying an alternative to using a major accident for this purpose.

The uncertainty analyses for the mechanistic source terms and r adiological doses should be described as part of the evaluation of conservative assumptions used in the DBA analysis. The plant design features intended to mitigate the radiological con sequences of accidents, the site atmospheric dispersion characteristics, and the distances to th e EAB and to the LPZ outer boundary are acceptable if the total calculated radiological co nsequences for the postulated fission product release meet the following reference values for public dose, given in 10 CFR 50.34(a)(1)(ii)(D) and 10 CFR 52.79(a)(1)(vi):

• An individual located at any point on the boundary of the exclu sion area for any 2-hour period following the onset of the postulated fission product re lease would not receive a radiation dose in excess of 25 rem TEDE, and;

• An individual located at any point on the outer boundary of the LPZ who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in ex cess of 25 rem TEDE.

10 For some non-LWR designs, assessment of the radiological consequences at the EAB and LPZ outer boundary using a major accident as specified in the regulations may be conservat ive or bounding and therefore establish compliance with 10 CFR 50.34 or 10 CFR 52.79. If assessment a major accident is not conservative or bounding for a specific non-LWR design in this regard or an applicant proposes to use a DBA that is not a major accident, justification for assessment of the radiological consequences at the EAB and LPZ outer boundary using a different event will be needed.

RG 1.253, Page 14 (2) Option 2: Use the greater of the dose consequence results from the bounding DBA and from a bounding BDBE, as identified in the LMP-based approach, to esta blish the acceptability of the EAB and LPZ. The uncertainty analyses for the mechanistic sourc e terms and radiological doses should be described as part of the evaluation of conservative a ssumptions used in the analysis.

This option provides an acceptable approach to compliance with 10 CFR 50.34 and 10 CFR 52.79 that precludes the need for an exemption from these requirement s, as long as the bounding BDBE involves or bounds an event sequence meeting the description of a major accident and the offsite consequences are below the reference values for public dose in 10 CFR 50.34(a)(1)(ii)(D)(1) or 10 CFR 52.79(a)(1)(vi)(A) for the EAB and those in 10 CFR 50.34 (a)(1)(ii)(D)(2) or 10 CFR 52.79(a)(1)(vi)(B) for the outer LPZ boundary.

c. Addition: Section C.2.1.1 of NEI 21-07, Revision 1, Overview o f PRA, includes a subsection titled, Two-Step Licensing (CP Content). This section notes that as p art of a CP application the applicant should address the last five items in the Section 2.1.1 list, c onsistent with the state of the plant design and the PRA at the time of the CP application. In addition to these five items, the application should include the item in the Section C.2.1.1 list labeled, Identifi cation of the sources of radionuclides addressed and the sources of radionuclides that were screened o ut. Commented [A17]: NRC-2022-0073-DRAFT-0006-22

d. Clarification: As noted in Section C.2.1.1 of NEI 21-07, a CP a pplicant should describe the attributes Commented [A18]: NRC-2022-0074-DRAFT-0006-22 of the PRA in the application. In addition to these attributes, as amended by position C.3.c above, the CP application should also discuss topics such as the PRAs con formance to RG 1.247 for trial use, and NEI 20-09, if a peer review is performed at the CP stage. Appendix A of this RG 1.253, Revision 0, provides additional guidance on demonstrating the a cceptability of the PRA supporting the CP application.

Commented [A19]: NRC-2022-0074-DRAFT-0006-23

e. Clarification: In addition to the site information described in Section C.2 of NEI 21-07, Revision 1, Commented [A20]: NRC-2022-0074-DRAFT-0006-24 the applicant in Chapter 2 of the SAR should provide informatio n not developed using the LMP process, including summaries of the site-related information an d analyses used to derive the design-basis hazard levels (DBHLs) documented in Chapter 6 of t he SAR. The purpose of this information is (1) to demonstrate compliance with 10 CFR Part 1 00, Reactor Site Criteria (Ref. 24),

Subpart B, Evaluation Factors for Stationary Power Reactor Sit e Applications on or after January 10, 1997, and the relevant site-related requirements o f 10 CFR Part 50 and 10 CFR Part 52, and (2) to describe the site characteristics used to inform the selection of the DBHLs in the design and safety analysis. Considerations for each relevant hazard includ e:

Commented [A21]: NRC-2022-0074-DRAFT-0006-24 (1) NSRST SSCs credited in non-DBA licensing basis events (i.e., AO Os, DBEs, and BDBEs) or to establish adequate DID may need to be specially designed to wit hstand or be protected from the hazard (e.g., application of special treatments in accordance w ith NEI 18-04 and RG 1.233), and

(2) NSRST SSCs relied upon to establish adequate DID for beyond-des ign-basis hazards may need to be designed with special treatment to withstand or be protected from each such hazard.

Commented [A22]: NRC-2022-0074-DRAFT-0006-24 The ARCAP ISG, DANU-ISG-2022-02, Site Information, (Ref. 25), contains additional guidance on one acceptable approach to determining the scope and level o f detail of the site information to be provided.

f. Addition: In addition to the information that NEI 21-07, Revisi on 1, states applicants should include in SAR Chapter 2, applicants should identify and describe the c ross-cutting engineering analyses and methodologies used to establish their design bases or to confir m that intended safety functions will be

RG 1.253, Page 15 fulfilled. Commented [A23]: NRC-2022-0074-DRAFT-0006-25

4. Licensing Basis Events

Section C.3 of NEI 21-07, Revision 1, provides guidance on the information related to the LBE selection methodology and the summary of LBEs (AOOs, DBEs, BDBE s, and DBAs) to include in an SAR. After identifying the LBEs, Chapter 3 of an SAR following NEI 21-07, Revision 1, should describe the systematic and reproducible process and methodology used to select the LBEs, and the specific analysis and evaluation of the selected LBEs for the proposed d esign. The analyses in this chapter are primarily described in terms of event sequences consisting of a n initiating event, the plant response to the initiating event (which includes a sequence of successes and fa ilures of mitigating systems), and a well-defined end state. Chapter 3 should also describe the proc ess used to group and condense the many event sequences considered in the PRA into event sequence famil ies that are used to define the AOOs, DBEs, and BDBEs. It is important to note that the term event s equence is used here, instead of the term accident sequence used in LWR PRA standards, because the scop e of the LBEs also includes AOOs and initiating events that do not result in radioisotope releas es.

It also important to note that for CP applicants, the requireme nts of 10 CFR 50.43(e)(1)(iii) to ensure that sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses do not apply. Accordingly, CP applicants are no t required to provide evaluations of the safety margins using approved evaluation models. However, preli minary analyses should be available to demonstrate the following:

1. The design will provide sufficient safety margins during normal operations and transient conditions.

2. The applicant has identified the SSCs necessary to prevent acci dents and mitigate accident consequences.

3. The applicant has demonstrated an understanding of the uncertai nty associated with the performance of SSCs necessary to prevent accidents and mitigate accident consequences.

The items above are closely related; for example, an understand ing of the uncertainties under item 3 is essential to an understanding of the margin under ite m 1. Additionally, items 2 and 3 support staff findings associated with 10 CFR 50.35(a)(3), namely, that the application describes the safety features and components that require research and development, and that the applicant will conduct a reasonably designed research and development program to resolve any associated safety questions (see the ARCAP Roadmap ISG on research and development).

C.4 Staff Position: NEI 21-07, Revision 1, Section C.3, on SAR Chapter 3, provide s an acceptable method for developing information related to the LBE selection methodology and the summary of LBEs (AOOs, DBEs, BDBEs, and DBAs), with the following clarifications.

a. Addition and Clarification: The discussion of AOOs, DBEs, and BDBEs in Chapter 3 of the SAR Commented [A24]: NRC-2022-0074-DRAFT-0006-26 should include a description of supporting data associated with the calculation of the mechanistic source terms and radiological consequences (to the extent that such information does not appear in the discussions of methodologies and analyses in Chapter 2, the des criptions of systems and functions in Chapters 5-7, or other sections of the SAR).

Commented [A25]: NRC-2022-0074-DRAFT-0006-26

RG 1.253, Page 16

b. Addition: In addition to the material identified in NEI 21-07, Revision 1, Section C.3, that is derived Commented [A26]: NRC-2022-0074-DRAFT-0006-28 by following the methodology of NEI 18-04, the applicant should address certain specified events and Commented [A27]: NRC-2022-0074-DRAFT-0006-29 requirements in a new SAR Section 3.7, Special Event Analyses, as described below: Commented [A28]: NRC-2022-0074-DRAFT-0006-30

(1) 10 CFR 50.150(b) requires that the preliminary safety analysis report (PSAR) or FSAR include a description of (a) the design features and functional capabilit ies identified in 10 CFR 50.150(a)(1)

(i.e., through the applicants assessment required by section 5 0.150(a)(1)), and (b) how the design features and functional capabilities identified in 10 CFR 50.15 0(a)(1) meet the assessment requirements in 10 CFR 50.150(a)(1). The ARCAP Roadmap ISG cont ains guidance regarding Commented [A29]: NRC-2022-0074-DRAFT-0006-31 aircraft impact assessments.

(2) Mitigation of Beyond-Design-Basis Events (MBDBE) (10 CFR 50.155, Mitigation of beyond-design-basis events (Ref. 26)): One of the primary less ons learned from the accident at the Fukushima Dai-ichi nuclear power plant in Japan was the sig nificance of the challenge presented by a loss of multiple SR systems after a beyond-desig n-basis external event. As a result of lessons learned from the Fukushima Dai-ichi accident, the NR C amended its regulations to establish requirements for nuclear power reactor applicants and licensees for mitigating beyond-design-basis events (i.e., 10 CFR 50.155(b)(1)).

In the case of the Fukushima Dai-ichi accident, the loss of all alternating current power led to loss of core cooling, and ultimately to core damage and a loss of co ntainment integrity. The design-basis for U.S. nuclear plants includes bounding analyses with m argin for external events expected at each site. Extreme external events (e.g., seismic events or external flooding, etc.) beyond those accounted for in the design-basis, while unlikely, could presen t challenges to nuclear power plants. The following documents provide guidance on implementat ion of the regulations at 10 CFR 50.155 and applicants using NEI 21-07, Revision 1, to de velop their applications should use the following documents:

• RG 1.226, Flexible Mitigation Strategies for Beyond-Design-Bas is Events (Ref. 27),

identifies methods and procedures the NRC staff considers accep table for nuclear power reactor applicants and licensees to use to demonstrate complian ce with NRC regulations on planning and preparedness for BDBEs as required by 10 CFR 50.15 5. RG 1.226 endorses, with clarifications, the methods and procedures in NEI 12-06, R evision 4, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, issued December 2016 (Ref. 28),

as a process the NRC considers acceptable for meeting, in part, the regulations in 10 CFR 50.155. Additionally, RG 1.226 provides guidance for mee ting the regulations in 10 CFR 50.155 in areas not covered by NEI 12-06, Revision 4.

• RG 1.227, Wide-Range Spent Fuel Pool Level Instrumentation (R ef. 29), identifies methods and procedures the NRC staff considers acceptable for d emonstrating compliance with NRC regulations on providing a reliable means to remotely monitor wide-range spent fuel pool levels to support implementation of event mitigation and recovery actions as required by 10 CFR 50.155. RG 1.227 endorses, with exceptions a nd clarifications, the methods and procedures in NEI 12-02, Revision 1, Industry Guid ance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliabl e Spent Fuel Pool Instrumentation, issued August 2012 (Ref. 30), as a process t he NRC staff considers acceptable for meeting certain regulations in 10 CFR 50.155.

• As noted in the statements of consideration for 10 CFR 50.155 ( 84 FR 39684) (Ref. 31), in recognition of the similarity of the existing extensive damage mitigation guidelines (EDMGs) formerly in 10 CFR 50.54(hh)(2) to the strategies required by 1 0 CFR 50.155(b)(1), the NRC

RG 1.253, Page 17 relocated the EDMGs into the MBDBE rule as 10 CFR 50.155(b)(2). The EDMGs provide strategies and guidelines to maintain or restore core cooling, containment, and spent fuel pool cooling capabilities under the circumstances associated with lo ss of large areas of the plant due to explosions or fire. The EDMGs provide strategies and gui delines in the following areas: firefighting, operations to mitigate fuel damage, and ac tions to minimize radiological release. NEI 06-12, B.5.b Phase 2 & 3 Submittal Guideline (Re f. 32), provides guidance on how to develop the application content for demonstrating that t he requirements of 10CFR 50.155(b)(2) are met.11

Note: Applicants for a CP that are not requesting design finali ty for mitigation of specific beyond design basis events reflected in 10 CFR 50.155(a) and applicant s for a DC are not required to provide information on this topic. Commented [A30]: NRC-2022-0074-DRAFT-0006-32

5. Integrated Evaluations

Section C.4 of NEI 21-07, Revision 1, provides guidance on docu menting the integrated evaluations performed using the LMP process in NEI 18-04, Revis ion 1. Chapter 4 of an SAR following NEI 21-07, Revision 1, should provide the overall plant risk pe rformance summary for the proposed design. This integrated plant evaluation assesses plant perform ance against the following three cumulative risk targets, and describes the margin between these targets an d the predicted plant performance:

• The total mean frequency of exceeding a site boundary dose of 1 00 millirem from all LBEs should not exceed 1/plant-year. The value of 100 millirem is taken fro m the annual cumulative exposure limits in 10 CFR Part 20, Standards for Protection against Rad iation (Ref. 33).

• The average individual risk of early fatality within 1 mile of the EAB from all LBEs, based on mean estimates of frequencies and consequences should not exceed 5x1 0-7/plant-year, to meet the NRC safety goal QHO for early fatality risk.

• The average individual risk of latent cancer fatalities within 10 miles of the EAB from all LBEs, based on mean estimates of frequencies and consequences should not exceed 2x10-6/plant-year, to meet the NRC safety goal QHO for latent cancer fatality risk.

Chapter 4 of an SAR following NEI 21-07, Revision 1, also docu ments the applicants assessment of the adequacy of DID for the plant design, address ing the three focus areas for DID adequacy: plant capability; programmatic capability; and integr ated risk-informed, performance-based DID adequacy. The baseline DID adequacy evaluation results in t his chapter and other SAR chapters should be documented in sufficient detail so that, before being implemented, proposed future changes to physical, functional, operational, or programmatic features of the facility can be effectively evaluated for their potential to reduce DID.

C.5 Staff Position: NEI 21-07, Revision 1, Section C.4, on SAR Chapter 4, provides an acceptable method for developing information related to the integrated evaluations, which include the overall plant risk performance summary, margins between predicted plant perfo rmance and risk targets, and the documentation of DID adequacy with the following clarifications and additions:

11 SRP Section 19.4, Strategies and Guidance to Address Loss-of-Large Areas of the Plant Due to Explosions and Fires provides guidance to the NRC staff for review of this topic. The SRP is intended to make information about regulatory matters widely available and to improve communication among the NRC, interested members of the public, and the nuclear power industry, thereby increasing understanding of the NRCs review process.

RG 1.253, Page 18

a. Clarification: The NRC anticipates that the DID discussion at t he CP stage may be limited to plant Commented [A31]: NRC-2022-0074-DRAFT-0006-33 capabilities because programmatic capabilities may not have bee n established yet. In addition, while not all plant capability DID attributes may be fully addressed at the CP stage, qualitative performance-based objectives for DID may be useful in establish ing performance boundaries for final safety analysis report results. The CP application should provi de a discussion in the SAR of the approach to establish DID adequacy. A discussion in the SAR to implement the DID adequacy Commented [A32]: NRC-2022-0074-DRAFT-0006-33 assessment processes in RG 1.233 is considered acceptable for t his purpose. Commented [A33]: NRC-2022-0074-DRAFT-0006-33

b. Addition: In addition to the results and margins, the SAR Chapt er 4 should include a summary of departures taken from or unique inputs to the methodologies des cribed in other chapters, if any, related to the analyses of cumulative risk measures. Examples t hat could arise due to factors such as the different time periods used in the assessments of licensing basis events and cumulative risk metrics could include sources of dose (cloud shine, inhalation, ground shine), additional inputs for dose conversion factors, and modeling assumptions (e.g., offsit e protective actions). The summary can be provided via references to other documents or guidance r elated to the assessment of cumulative risk metrics.

Commented [A34]: NRC-2022-0074-DRAFT-0006-34

c. Clarification: Human factors considerations for SSCs should be included in SAR Chapter 6 or 7, as NRC-2022-0075-DRAFT-0004-4 appropriate. The human factors information in these SAR chapter s should be consistent with the human factors information provided in SAR Chapter 11 in accorda nce with ARCAP DANU-ISG-2022-05, Organization and Human Systems Considerations. (Ref. 34). Commented [A35]: NRC-2022-0074-DRAFT-0006-35

d. Clarification: Guidance for the change control process for the SAR, including ensuring the design and construction of defense-in-depth features remains adequate (i.e., up to issuance of an operating license), is addressed in NEI 18-04, Revision 1 as endorsed by RG 1.233. Additional guidance related to change control for the FSAR following issuance of the operat ing license is under development and the NRC is not taking a position on this topic at this time. Th e staff may address such change control processes in future regulatory actions, including possible rule makings, license conditions, and development of guidance documents. Commented [A36]: NRC-2022-0074-DRAFT-0006-36 NRC-2022-0075-DRAFT-0004-5

6. Safety Functions, Design Criteria, and SSC Safety Classifications

Section C.5 of NEI 21-07, Revision 1, provides guidance on the information related to safety functions, design criteria, and SSC classification established using the LMP process in NEI 18-04, Revision 1, and endorsed in RG 1.233. In the LMP process, LBEs are generally defined in terms of successes and failures of SSCs that perform safety functions an d are modeled in the PRA. Therefore, the PRA safety functions (PSFs) are those functions credited for pr eventing or mitigating unplanned radiological releases from any source within the plant.

Chapter 5 of an SAR following NEI 21-07, Revision 1, should des cribe the applicants approach to designating SSC safety functions and classifications in acco rdance with the PSFs. For SSCs, the applicant should describe the required safety functions (RSFs), the required functional design criteria (RFDC), the PDC, and the classification of SR and NSRST SSCs. These terms are defined in NEI 21-07, Revision 1, Appendix A, Glossary of Terms.

C.6 Staff Position: NEI 21-07, Revision 1, Section C.5, provides an acceptable me thod for developing information related to the safety classification of SSCs, including information about RSFs,

RFDC, PDC, and SR and NSRST SSCs with the following clarificati ons and additions:

a. NEI 21-07, Revision 1, Chapter 5, provides an acceptable approa ch for developing proposed PDC, with the following clarifications. Commented [A37]: NRC-2022-0074-DRAFT-0006-39

RG 1.253, Page 19 (1) Clarification: The inclusion of a proposed quality assurance PD C as described in Chapter 5 of Commented [A38]: NRC-2022-0074-DRAFT-0006-37 NEI 21-07, Revision 1, is an acceptable method for implementing a graded approach to quality assurance for SSCs; it can also contribute to the basis for not addressing quality assurance in the scope of PDC in the more system-and component-specific PDC pro posed.

(2) Clarification: As described in NEI 18-04, Revision 1, and RG 1. 233, a non-LWR applicant may use a risk-informed methodology (e.g., the LMP methodology) to identify both RSFs and PSFs from which to determine RFDC and other special treatment requir ements for SR and NSRST SSCs. The role of the RFDC and special treatment requirements d erived from the LMP process in identifying design features and related attributes is similar t o that of the advanced reactor design criteria and the requirements of the GDC. Therefore, to meet th e regulations for proposing PDC, the scope of the proposed PDC should include SSCs important to safety. For applicants using the LMP process endorsed in RG 1.233, SSCs important to safety incl ude both SR and NSRST SSCs.

Therefore, the proposed PDC will need to address the functions provided by both SR and NSRST SSCs. NEI 21-07, Revision 1, Chapter 5, describes a two-tiered approach to PDC, comprising a higher-level portion based on meeting functional design criteri a through RFDC and a bottom-up portion based on meeting specific performance requirements thro ugh complementary design criteria (CDC). This two-tiered approach proposed in NEI 21-07, Revision 1, divides PDC into PDC-RFDC and PDC-CDC. Commented [A39]: NRC-2022-0074-DRAFT-0006-38

(3) Clarification: Applicants adopting alternative approaches to pr oposing PDC based on similar Commented [A40]: NRC-2022-0074-DRAFT-0006-39 risk-informed, performance-based licensing methodologies should provide suitable justification for their approaches and include any exemptions necessary. Exem ptions from the regulations addressing content of applications are necessary if the full sc ope of PDC, as discussed above, is not addressedthat is, if the PDC do not cover all necessary de sign, fabrication, construction, testing, and performance requirements for all SSCs important to safety. For example, the justification may be that, to address specific elements of PDC scope not included here, the applicant has complied with other regulatory requirements that compel the applicant to provide the relevant information in other portions of the application.

b. Addition: Additional information on the role of fuel: In additi on to the material identified in Commented [A41]: NRC-2022-0074-DRAFT-0006-40 NEI 21-07, Revision 1, Section C.5, Chapter 5 of an SAR followi ng NEI 21-07, Revision 1, should NRC-2022-0075-DRAFT-0004-7 also address fuel qualification. The reactor core and its fuel are generally classified as SR because they are directly involved in performing fundamental safety fun ctions. The application should provide the information for SR SSCs identified in NEI 21-07, Revision 1, Chapter 6, Safety-Related SSC Criteria and Capabilities. However, the adequacy of fuel perfo rmance also depends on other information such as fuel design limits and fuel qualification, which the application should describe. Commented [A42]: NRC-2022-0074-DRAFT-0006-40 The applicants discussion should focus on the role of the fuel in the safety analysis for the reactor NRC-2022-0075-DRAFT-0004-7 and on the adequacy of the plan to provide the basis for fuel p erformance as credited in the safety analysis. If not included elsewhere in the application or in re ferenced reports, this section of the SAR should include information sufficient to establish that:

(1) The role of the fuel in the safety analysis is adequately descr ibed. This can be accomplished by stating how the fuel will perform during (a) normal operation, including the effects of AOOs, and (b) accident conditions. To support these findings, sufficient information should be provided to clearly identify the design limits of the fuel and the fuel con tribution in the accident source term.

The applicants discussion of the design limits and source term should address uncertainty from any limitations on data available, as reflected in the analyses discussed in Chapters 2 and 3 of NEI 21-07, Revision 1.

RG 1.253, Page 20 (2) The fuel qualification plan is adequate. The discussion of the fuel qualification plan should consider the proposed analysis methodologies (e.g., fuel perfor mance codes), the use of existing data, and any ongoing testing or plans to use lead test specime ns. If the applicant is using legacy data, it should justify the applicability of the data to the pr oposed facility (e.g., by confirming that the data were collected for a fuel fabricated consistent with t he proposed fuel design and irradiated in an appropriate environment among other factors).

7. Safety-Related (SR) Systems, Structures, and Components Criteria and Capabilities

Section C.6 of NEI 21-07, Revision 1, provides guidance on the information related to the SR classification of SSCs, as well as their associated design crit eria and performance capabilities. Chapter 6 of an SAR following NEI 21-07, Revision 1, should give details on SSCs classified as SR following the guidance in Chapter 5 of NEI 21-07, Revision 1. In particular, the SAR should give further detail on all design criteria and performance capabilities applying to SR SSC s, including safety-related design criteria, performance-based targets for reliability and capabilities, and special treatment requirements to provide sufficient confidence that the performance-based targets for th e design will be achieved in the construction of the plant and maintained throughout the license d plant life. For those SR SSCs whose reliability and capabilities have not been confirmed at the CP stage, the PSAR should include sufficient information (e.g., commitments for testing or research and deve lopment) to confirm that the reliability and capability performance targets informed by the final PRA wi ll be met. Commented [A43]: NRC-2022-0074-DRAFT-0006-41 Commented [A44]: NRC-2022-0074-DRAFT-0006-42 The term special treatment is derived from NRC regulations an d NEI guidelines for implementing 10 CFR 50.69, Risk-informed categorization and tr eatment of structures, systems and components for nuclear power reactors. RG 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safet y Significance (Ref. 35), defines special treatment as those requirements that provide increased assurance beyond normal industrial practices that structures, systems, and components (SSCs) perfo rm their design-basis functions.

Special treatments are considered anything that is done, beyond procuring commercial-grade equipment, to provide increased assurance of the capability and reliability of both SR and NSRST SSCs, including, for example, design requirements, quality assurance requirements, availability controls, reliability and capability controls, and monitoring programs as sociated with reliability. Table 4-1 of NEI 18-04, Revision 1, gives additional information on possible types of special treatments that may be considered for an SSC. Chapter 6 of an SAR following NEI 21-07, Revision 1, should include information on the special treatments selected for SR SSCs.

One category of design requirements for SR SSCs consists of tho se measures or requirements needed to protect them from or ensure their ability to withstan d the adverse effects of design-basis hazards when performing their RSFs. These design-basis hazards include both internal and external hazards; they are characterized as DBHLs that SR SSCs must have the ability to withstand or from which SR SSCs must be protected. DBHLs may be selected either determi nistically or probabilistically.

Chapter 6 of the SAR provides information on the establishment of the applicable DBHLs, the bases for establishment, and the associated parameters that lead to desig n requirements for SR SSCs.

C.7 Staff Position: NEI 21-07, Revision 1, Section C.6, on SAR Chapter 6, provide s an acceptable method for developing information related to SSC design require ments and capabilities, including DBHLs, special treatment requirements, and system descriptions for SR SSCs with the following clarifications and additions:

RG 1.253, Page 21

a. Addition: In addition to describing the DBHLs as stated in NEI 21-07, Revision 1, Section C.6, the applicant may also use the guidance in section C.I.3 of RG 1.20 6, Revision 0 (Ref. 42),12 to determine the information that should be included in Chapters 5 and 6 of the SAR regarding the translation of DBHLs to loads on SSCs, evaluation of those loads, and related design analysis. Pre-application Commented [A45]: NRC-2022-0074-DRAFT-0006-45 interactions with the staff may be appropriate to determine the necessary level of information to be NRC-2022-0074-DRAFT-0006-46 included in the SAR. NRC-2022-0075-DRAFT-0004-11

8. Non-Safety-Related with Special Treatment (NSRST) Structures, Systems, and Components Criteria and Capabilities

Section C.7 of NEI 21-07, Revision 1, provides guidance on the information related to the NSRST classification of SSCs, as well as their associated crite ria and capabilities. Chapter 7 of an SAR following NEI 21-07, Revision 1, should describe the design and special treatment requirements for those SSCs classified as NSRST in Chapter 5 of the SAR. NSRST SSCs are not directly associated with RFDC (i.e., they are not SR SSCs) but are relied upon to perform risk-significant functions. Special treatments are defined above, with additional information provided in NEI 18-04, Revision 1. Risk-significant SSCs are those that perform functions that prevent any LBE from exce eding the frequency-consequence targets or that contribute significantly to the cumulative risk metrics selected for evaluating the total risk from all analyzed LBEs. Appendix A to NEI 21-07, Revision 1, gives a mor e detailed definition of risk-significant SSCs. NSRST reliability and capability targets can be provided at the CP or the OL stage. For those NSRST SSCs whose reliability and capabilities have not been pro vided and confirmed at the CP stage, the application should include a discussion in the PSAR on how the applicant intends to confirm, at the OL stage, that reasonable reliability and capability performanc e targets have been established, align with the supporting analyses, and have special treatments defined to ensure the performance of SSCs meet the targets. The OL application should describe any testing and val idation confirming NSRST SSC performance capabilities and availability, including any additi onal special treatments to be applied to the NSRST SSCs as compensatory measures to address a lack of operat ing experience. Commented [A46]: NRC-2022-0074-DRAFT-0006-50 NRC-2022-0075-DRAFT-0004-3 C.8 Staff Position: NEI 21-07, Revision 1, Section C.7, provides an acceptable met hod for NRC-2022-0073-DRAFT-0011-3 developing information related to the special treatment require ments for NSRST SSCs and the descriptions and capabilities of NSRST SSCs. Table 4-1 of NEI 1 8-04, Revision 1, gives additional information on the types of special treatments that may be cons idered for SSCs.

9. Plant Programs

Section C.8 of NEI 21-07, Revision 1, provides guidance on the information related to plant programs that support the LMP-based safety analysis. Chapter 8 of an SAR following NEI 21-07, Revision 1, should give an overview of the plant programs relie d upon to support the LMP-based safety analysis, addressing these programs purpose, scope, and perfor mance objectives, as well as applicability to SR SSCs, NSRST SSCs, and operations activities. The applican t should describe the performance objectives of each program and explain how they relate to the t argets or special treatments identified for SR and NSRST SSCs. This information should be included in the S AR or in documents that are incorporated by reference. Construction permit applications sho uld include general descriptions in the SAR regarding any programs needed to implement special treatmen ts and meet reliability and Commented [A47]: NRC-2022-0074-DRAFT-0006-53 performance targets for SR SSCs and NSRST SSCs. These may inclu de programs for inservice inspection/testing, maintenance, human factors, training, and r eliability assurance.

12 Chapter 3 of NUREG-0800 provides guidance to the NRC staff for review of this topic. The SRP is intended to make information about regulatory matters widely available and to im prove communication among the NRC, interested members of the public, and the nuclear p ower industry, thereby increasing understanding of the NRCs review process.

RG 1.253, Page 22 Chapter 8 should cover those plant programs used for special tr eatments for SR and NSRST SSCs (as described in Chapters 6 and 7, respectively) to ensure that (1) reliability and performance targets are met, and (2) safety-significant uncertainties are addressed as part of DID. In addition, Chapter 8 should also identify and give an overview of the program or programs f or documenting SSC reliability and capability targets, as described in Chapters 6 and 7 and ensuri ng that these targets are met. Program areas could also include human factors, quality assurance, startup te sting, and equipment qualification, among others. The discussion of plant programs should address the dif ferent plant lifetime phases (i.e., design, construction, testing, and operations), as applicable.

C.9 Staff Position: NEI 21-07, Revision 1, Chapter 8, provides an acceptable meth od for developing information related to plant programs relied upon to support th e LMP-based safety analyses, including programs used to implement special treatments for SR and NSRST SSCs and to meet reliability and capability targets.

RG 1.253, Page 23 D. IMPLEMENTATION

The NRC staff may use this regulatory guide as a reference in i ts regulatory processes, such as licensing, inspection, or enforcement. However, the NRC staff d oes 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 50.109, Backfitting, and as described in NR C Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, an d Information Requests (Ref. 36), nor does the NRC staff intend to use the guidance to affect the iss ue finality of an approval under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nu clear Power Plants. 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. If a licensee believes that the NRC is using this RG in a manner inconsistent with the discussion in t his 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.

RG 1.253, Page 24 ACRONYMS/ABBREVIATIONS

ACI American Concrete Institute ADAMS Agencywide Documents Access and Management System ANS American Nuclear Society AOO anticipated operational occurrence ARCAP advanced reactor content of application project ASME American Society of Mechanical Engineers BDBE beyond-design-basis event CDC complementary design criterion/a CFR Code of Federal Regulations COL combined license CP construction permit DC design certification DBA design-basis accident DBE design-basis event DBHL design-basis hazard level DG draft regulatory guide DID defense in depth EAB exclusion area boundary FSAR final safety analysis report IAEA International Atomic Energy Agency ISG interim staff guidance LBE licensing-basis event LMP Licensing Modernization Project LPZ low-population zone LWR light-water reactor ML manufacturing license NEI Nuclear Energy Institute NEIMA Nuclear Energy Innovation and Modernization Act NRC U.S. Nuclear Regulatory Commission NSRST non-safety-related with special treatment NST non-safety-related with no special treatment OL operating license OMB Office of Management and Budget PDC principal design criterion/a PRA probabilistic risk assessment PSAR preliminary safety analysis report PSF PRA safety function QHO quantitative health objective RFDC required functional design criterion/a RG regulatory guide RSF required safety function SAR safety analysis report SDA standard design approval SR safety-related SSC structure, system, or component TEDE total effective dose equivalent TICAP technology-inclusive content of application project U.S.C. United States Code

RG 1.253, Page 25 REFERENCES 13

1. Nuclear Energy Institute (NEI), NEI 21-07, Revision 1, Technol ogy Inclusive Guidance for Non-Light Water Reactors, Safety Analysis Report Content: For A pplicants Using the NEI 18-04 Methodology, Washington, DC, February 2022. (Agencywide Docume nts Access and Management System (ADAMS) Accession No. ML22060A190)

2. Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Domestic Licensing of Production and Utilization Facilities.

3. 10 CFR Part 52, Licenses, Certifications, and Approvals f or Nuclear Power Plants.

4. NRC, Policy Statement on the Regulation of Advanced Reacto rs, Federal Register, Vol. 73, No. 199, October 14, 2008, pp. 60612-60616 (73 FR 60612).

5. NRC, RG 1.70, Standard Format and Content of Safety Analys is Reports for Nuclear Power Plants (LWR Edition), Washington, DC.

6. NRC, RG 1.181, Content of the Updated Final Safety Analysi s Report in Accordance with 10 CFR 50.71(e), Washington, DC.

7. NRC, RG 1.206, Applications for Nuclear Power Plants, Was hington, DC.

8. NRC, RG 1.232, Guidance for Developing Principal Design Cr iteria for Non-Light-Water Reactors, Washington, DC.

9. RG 1.233, Guidance for a Technology-Inclusive, Risk-Inform ed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applic ations for Licenses, Certifications, and Approvals for Non-Light-Water Reactors, Wa shington, DC.

10. NEI 18-04, Revision 1, Risk-Informed Performance-Based Te chnology-Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Washingto n DC, August 2019.

(ML19241A472)

13 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 (PDR) staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC 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.

Publications from the Nuclear Energy Institute (NEI) are avail able at their Web site: http://www.nei.org/ or by contacting the headquarters at Nuclear Ene rgy Institute, 1776 I Street NW, Washington DC 20006-3708; telephone: 202-739-800; fax 202-785-4019.

Copies of International Atomic Energy Agency (IAEA) documents may be obtained through its Web site: www.iaea.org or by writing the International Atomic Energy Agency, P.O. Box 100, Wagramer Strasse 5, A-1400 Vienna, Austria.

RG 1.253, Page 26

11. NRC, NRC Vision and Strategy: Safely Achieving Effective and Efficient Non-Light Water Reactor Mission Readiness, December 2016. (ADAMS Accession No. ML16356A670)

12. NRC, NRC Non-Light Water Reactor Near-Term Implementation Action Plans, July 2017.

(ML17165A069)

13. NRC, NRC Non-Light Water Reactor Mid-Term and Long-Term I mplementation Action Plans, July 2017. (ML17164A173)

14. NRC, SRM-COMGBJ-10-0004/COMGEA-10-0001, Staff Requirement s COMGBJ-10-0004/COMGEA-10-0001Use of Risk Insights to Enhance S afety Focus of Small Modular Reactor Reviews, August 31, 2010. (ML102510405)

15. NRC, SRM-SECY-11-0024, Staff RequirementsSECY-11-0024Us e of Risk Insights to Enhance the Safety Focus of Small Modular Reactor Reviews, May 11, 2011. (ML111320551)

16. NRC, DANU-ISG-2022-01, Review of Risk-Informed, Technolo gy-Inclusive Advanced Reactor Applications-Roadmap, Washington, DC (ML23277A139)

17. NRC, NUREG 0800, Standard Review Plan for the Review of S afety Analysis Reports for Nuclear Power Plants: LWR Edition, Washington, DC

18. NRC, Nuclear Regulatory Commission International Policy S tatement, Federal Register, Vol. 79, No. 132, July 10, 2014, pp. 39415-39418 (79 FR 39415).

19. NRC, Management Directive (MD) 6.6, Regulatory Guides, W ashington, DC.

20. International Atomic Energy Agency (IAEA) Specific Safety Requirements (SSR), No. SSR-2/1, Safety of Nuclear Power Plants: Design, Vienna, Austria, 2016.

21. NRC, RG 1.247 for trial use, Acceptability of Probabilist ic Risk Assessment Results for Non-Light-Water Reactor Risk-Informed Activities, Washington, DC.

22. American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS)

RA-S-1.4-2021, Probabilistic Risk Assessment Standard for Adva nced Non-Light Water Reactor Nuclear Power Plants, New York, NY, 2021.

23. NEI 20-09, Performance of PRA Peer Reviews Using the ASME /ANS Advanced Non-LWR PRA Standard, Washington DC, May 2021. (ML21125A284)

24. 10 CFR Part 100, Reactor Site Criteria.

25. NRC, DANU-ISG-2022-02, Site Information, Washington, DC. (ML23277A140)

26. 10 CFR 50.155, Mitigation of beyond design-basis events.

27. NRC, RG 1.226, Revision 0, Flexible Mitigation Strategies for Beyond-Design-Basis Events, Washington, DC, June 2019. (ML19058A012)

RG 1.253, Page 27

28. NEI 12-06, Revision 4, Diverse and Flexible Coping Strat egies (FLEX) Implementation Guide, Washington, DC, December 2016. (ML16354B421)

29. NRC, RG 1.227, Revision 0, Wide-Range Spent Fuel Pool Le vel Instrumentation, Washington, DC, June 2019. (ML19058A013)

30. NEI 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, Washington, DC, August 2012. (ML12240A307)

31. NRC, Final Rule, Mitigation of Beyond-Design-Basis Event s, Federal Register, Vol. 84, No. 154 August 9, 2019, pp. 39684-39722 (84 FR 39684).

32. NEI 06-12, Revision 2, B.5.b Phase 2 & 3 Submittal Guide line, Washington DC, December 2006.

33. 10 CFR Part 20, Standards for Protection against Radiati on.

34. NRC, DANU-ISG-2022-05, Organization and Human System Con siderations, Washington, DC, (ML23277A143)

35. NRC, RG 1.201, Guidelines for Categorizing Structures, S ystems, and Components in Nuclear Power Plants According to Their Safety Significance, Washingto n, DC, May 2006.

(ML061090627)

36. NRC, MD 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, Washington, D.C.

RG 1.253, Page 28 Appendix A Commented [A48]: NRC-2022-0074-DRAFT-0006-17 NRC-2022-0074-DRAFT-0006-38 NRC-2022-0074-DRAFT-0006-39 Acceptability of a Probabilistic Risk Assessment That Supports a Non-Light-Water Reactor Construction Permit Application Based on the Lice nsing Modernization Project Methodology

A.1 Introduction

This appendix provides supplemental guidance on one approach th at is acceptable to the U.S.

Nuclear Regulatory Commission (NRC) staff for preparing a proba bilistic risk assessment (PRA) for a non-light-water reactor (non-LWR) construction permit (CP) appl ication (also referred to as a CP PRA1) under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. A-1), based on the Licensing Modernization Project (LMP) methodology in the Nuclear Energy Institute (NEI) report NEI 1804, Revisio n 1, Risk-Informed Performance-Based Guidance for Non-Light-Water Reactor Licensing Basis Developmen t (Ref. A-2). The NRC staff developed this guidance with the goal of informing all stakehol ders, including applicants, of the type and detail of PRA information to be included in a non-LWR CP applic ation that would be sufficient to provide confidence in the PRA results such that the PRA can be used in regulatory decision making. Accordingly, the application should demonstrate that: Commented [A49]: NRC-2022-0073-DRAFT-0013-1 NRC-2022-0073-DRAFT-0013-2

• Commensurate with the preliminary plant design and proposed sit e described in the CP NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 application, information developed from the PRA is sound and re liable. NRC-2022-0073-DRAFT-0013-10 NRC-2022-0073-DRAFT-0013-11

• The PRA produces insights with appropriate fidelity to support implementation of the LMP methodology and development of the CP application.

• The CP applicant has defined processes and procedures to adequa tely maintain and upgrade the PRA to support continued implementation of the LMP methodology as the detailed plant design evolves and the plant is constructed, leading to submittal of t he operating license (OL) application.

The term PRA acceptability describes the ability of a PRA to support risk-informed regulatory decision making and is defined in terms of meeting the NRC regu latory positions in Section C of Regulatory Guide (RG) 1.247 (for trial use), Acceptability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor Risk-Informed Activities (Ref. A-3). Specifically, Regulatory Position C.1 of RG 1.247 and its subsections provide guidance in the following four areas that are collectively assessed to determine the acceptability of a PRA:

1. Scope of a PRA: The scope of a PRA is defined in terms of (1) t he metrics used to characterize risk, (2) the radiological sources that may contribute to risk, (3) the causes of initiating events (hazard groups) that can potentially challenge and disrupt the normal operation of the plant and, if not prevented or mitigated, would eventually result in a radioa ctive release, and (4) the plant

1 In this Appendix, a CP PRA refers to the collection of analyses represented by PRA documentation submitted as part of a CP application and the PRA documentation maintained by the applicant, which will be made available to the NRC staff through regulatory audits or in response to requests for additional information. Guidance on the four aspects of an acceptable CP PRA is provided in Sections A.3 through A.6 of this Appendix. Additionally, for the Technology-Inclusive Content of Application Project, the PRA includes the collection of analyses that represent risk contributors that are included in or screened out of the PRA logic model or that are accounted for by a risk-informed supplementary evaluation.

RG 1.253, App. A, Page A-1 operating states (POSs) for which the risk is to be evaluated. The scope of a CP PRA is determined by its intended uses for representing the as-designe d, as-to-be-built, and as-to-be-operated plant.

2. Level of detail of a PRA: The level of detail of a CP PRA is de fined in terms of the resolution of the modeling used to represent the behavior and operations of t he plant. A minimum level of detail is necessary to ensure that the impacts of designed-in d ependencies (e.g., support system dependencies, functional dependencies, and dependencies on oper ator actions) are correctly represented. This minimum level of detail is implicit in the el ements comprising the CP PRA and their associated characteristics and attributes.

3. Elements of a PRA: The PRA elements are defined in terms of the fundamental technical analyses used to develop and quantify the CP PRA model for its intended purpose (e.g., determination of a specific risk metric). The characteristics and attributes of th e PRA elements define specific criteria for successfully performing those technical analyses a nd achieving a defined objective.

4. Plant representation and PRA configuration control: Plant repre sentation is defined in terms of how closely the CP PRA represents the plant as it is designed, built, and operated. In general, CP PRA results should be derived from a CP PRA model that represen ts the as-designed, as-to-be-built, and as-to-be-operated plant. Consequently, the CP PRA sh ould be developed using an acceptable configuration control process.

The following sections provide guidance on these four interrela ted areas of PRA acceptability and PRA documentation in the context of a CP application that is ba sed on the LMP methodology. Consistent with NEI 1804, Revision 1, and Section C.2.1.1 of NEI 2107, R evision 1, Technology-Inclusive Commented [A50]: NRC-2022-0073-DRAFT-0013-1 Guidance for Non-Light Water Reactors, Safety Analysis Report C ontent: For Applicants Using the NRC-2022-0073-DRAFT-0013-2 NEI 18-04 Methodology (Ref. A-4), and their endorsements in RG 1.233, Guidance for a Technology-NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and NRC-2022-0073-DRAFT-0013-10 Content of Applications for Licenses, Certifications, and Appro vals for Non-Light-Water Reactors NRC-2022-0073-DRAFT-0013-11 (Ref. A-5), and the main body of this RG, this appendix assumes that the CP applicant will use the American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) non-LWR PRA standard, ASME/ANS RA-S-1.4-2021, Probabilistic Risk Asses sment Standard for Advanced Non-Light Water Reactor Nuclear Power Plants (Ref. A-6), to demons trate the acceptability of the PRA. CP applicants may use other approaches to demonstrate the acceptab ility of the PRA with appropriate justification; the NRC staff will review such an approach on a case-by-case basis. An applicant should maintain more detailed PRA information that supports or confirms the acceptability of the PRA in plant records (i.e., archival documentation) that are available for a udit by the NRC staff.

A.2 General

A.2.1 As discussed in Section B of Appendix A to DANU-ISG-2022- 01, Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications-Roadmap (Re f. A-7), prospective CP applicants may find it beneficial to engage in pre-application activities with the NRC staff regarding approaches to demonstrating the acceptability of a CP PRA before the LMP-base d CP application is submitted.

RG 1.253, App. A, Page A-2 A.2.2 Consistent with NEI 2107, Revision 1, Section 2.1.1, th e CP applicant should clearly document in the PSAR the essential assumptions 2 made in developing the LMP-based safety analysis, which should include those assumptions relevant to the probability and conse quence models, and the selection of elements to be incorporated in the CP PRA models. Commented [A51]: NRC-2022-0073-DRAFT-0013-3

A.2.3 The CP applicant should consider the near-term and long-t erm uses of the PRA as the PRA is developed to help ensure that it will be acceptable to support these uses. In addition to supporting implementation of the LMP methodology, results of the PRA may b e used to demonstrate how certain regulations and Commission policies have been met and to suppor t voluntary risk-informed applications, as discussed below.

Demonstrating that Certain Regulations Are Met

Currently, no regulation requires the development of a PRA to support a CP application under 10 CFR Part 50.3 However, the CP applicant may use the PRA to demonstrate, in p art, that the following regulations in 10 CFR Part 50 have been met:

1. 10 CFR 50.34(a)(1)(ii), which states, It is expected that reactors will reflect through their design, construction and operation an extremely low probability for acc idents that could result in the release of significant quantities of radioactive fission produc ts.

2. 10 CFR 50.34(a)(4), which requires the PSAR to include, [a] pr eliminary analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and s afety resulting from operation of the facility and including determination of the margins of safety d uring normal operations and transient conditions anticipated during the life of the facilit y, and the adequacy of structures, systems, and components provided for the prevention of accident s and the mitigation of the consequences of accidents.

Commission Policy Positions

The Commissions Policy Statement on the Regulation of Advanc ed Reactors (Volume 73 of the Federal Register (FR), page 60612 (73 FR 60612); October 14, 2008) (Ref. A-10) cites the following policy statements that express Commission expectations for use of the PRA. Specifically:

1. The advanced reactor policy statement articulates the expectati on that advanced reactor designs will comply with the Commissions safety goal policy statement (Safety Goals for Operations of Nuclear Power Plants; Policy Statement; Republication, 51 FR 2 8044; August 4, 1986, as corrected and republished at 51 FR 30028; August 21, 1986) (Ref. A-11). The safety goal policy

2 The term key assumption is similar to the essential assumptions referred to in Section C.2.1.1 of NEI 21-07, Revision 1. A key assumption is defined in NUREG-2122 Glossary of Risk-Related Terms in Support of Risk-Informed Decisionmaking (Rev. A-8) as [a]n assumption is considered to be key to a ris k-informed decision when it could affect the PRA results that are being used in a decision and, consequently, may influence the decision being made.

3 An ongoing rulemaking effort, Incorporation of Lessons Learned from New Reactor Licensing Process (10 CFR Parts 50 and 52 Licensing Process Alignment), Docket NRC-2009-0196, RIN-3150-AI66, includes proposed PRA-related requirements for 10 CFR Part 50 CP and OL applications that are similar to the existing PRA-related requirements for 10 CFR Part 52 licenses, certifications, and approvals. Further information about this rulemaking (including the proposed schedule) is available at https://www.nrc.gov/reading-rm/doc-collections/rulemaking-ruleforum/active/ruledetails.html?id=27.

RG 1.253, App. A, Page A-3 statement broadly defines an acceptable level of radiological r isk and establishes two qualitative safety goals which are supported by two quantitative objectives. Consistent with the safety goal policy statement, PRA is an acceptable tool for assessing confo rmance with the underlying purposes of the safety goals.

2. The advanced reactor policy statement notes that the Commission has issued a policy statement on Severe Reactor Accidents Regarding Future Designs and Exist ing Plants (50 FR 32138; August 8, 1985) (Ref. A-12), which indicates, in the context of the decision process for certifying a new standard plant design, that a new design can be shown to be acceptable for severe accident concerns, in part, by completion of a PRA and consideration of the severe accident vulnerabilities the PRA exposes along with the insights it may add to the assur ance of no undue risk to public health and safety.

3. The advanced reactor policy statement indicates the use of PRA as a design tool is implied by the Commissions policy statement on Use of Probabilistic Risk Ass essment Methods in Nuclear Regulatory Activities (60 FR 42622; August 16, 1995) (Ref. A-1 3).

Implementation of the LMP methodology inherently conforms to t he underlying purposes of these Commission policies.

Supporting Risk-Informed Applications of the PRA in Addition to an Initial Licensing Application

CP applicants may use the PRA to support risk-informed applica tions, in addition to implementing the LMP methodology, either concurrently with the CP application, concurrently with the OL application, or after the OL issuance. These additional risk -informed applications may affect the PRA scope, level of detail, or elements that should be considered e arly during PRA development. Examples of additional risk-informed applications include, but are not limi ted to:

1. Risk-informed inservice inspection and inservice testing progra ms. Guidance is provided in DANU-ISG-2022-07, Risk-Informed Inservice Inspection/Inservice Testing Programs for Non-LWRs (Ref. A-14).

2. Programs to implement the 2019 Edition of the ASME Boiler and P ressure Vessel Code (ASME Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants (Ref. A-15). Guidance is provided in RG 1.246, A cceptability of ASME Code,Section XI, Division 2, Requirements for Reliability and Integ rity Management (RIM) Programs for Nuclear Power Plants (Ref. A-16).

3. Risk-informed technical specifications. Guidance is provided in DANU-ISG-2022-08, Risk-Informed Technical Specifications (Ref. A-17).

4. Risk-informed fire protection programs. Guidance is provided in DANU-ISG-2022-09, Risk-Informed, Performance-Based Fire Protection Program ( for Operations) (Ref. A-18).

A.3 PRA Scope

A.3.1 A CP applicant using the LMP methodology should estimate the risk metrics described in Commented [A52]: NRC-2022-0073-DRAFT-0013-4 Sections C.3.2.1, C.4.1.1, C.4.1.2, and C.4.1.3 of NEI 21-07, R evision 1, on either a qualitative or quantitative basis consistent with the information available wh en the CP application is prepared. Such an

RG 1.253, App. A, Page A-4 estimation should include an explanation of how the Commission s quantitative health objectives (QHOs) from the Commission Safety Goal Policy Statement will be met in support of the OL application. Commented [A53]: NRC-2022-0073-DRAFT-0013-1 NRC-2022-0073-DRAFT-0013-2 A.3.2 Consistent with NEI 2107, Revision 1, Section 2.1.1, as clarified in Staff Position C.3.c in the NRC-2022-0073-DRAFT-0013-5 main body of this regulatory guidance and RG 1.247 (for trial u se), Staff Position C.1.1, the CP applicant NRC-2022-0073-DRAFT-0013-7 NRC-2022-0073-DRAFT-0013-10 should: NRC-2022-0073-DRAFT-0013-11 Commented [A54]: NRC-2022-0073-DRAFT-0013-5

1. Identify all radiological sources, hazards, and POSs by performing a comprehensive and systematic search.

2. Disposition the search results by a combination of PRA logic mo deling, acceptable screening methods, risk-informed supplemental evaluations, and crediting design-basis hazard levels (DBHLs).

A.3.3 Regarding PRA acceptability, the minimum scope of the CP PRA logic model should include the Commented [A55]: NRC-2022-0073-DRAFT-0013-1 internal events hazard group for the reactor in the at-power PO S. At a minimum, this demonstrates the NRC-2022-0073-DRAFT-0013-2 applicants ability to develop an acceptable licensing basis an d to establish an acceptable foundation for NRC-2022-0073-DRAFT-0013-5 upgrading the PRA logic model as the design progresses. NRC-2022-0073-DRAFT-0013-7 NRC-2022-0073-DRAFT-0013-10 NRC-2022-0073-DRAFT-0013-11 A.3.4 Regarding PRA acceptability, the high-level requirements 4 and associated supporting Commented [A56]: NRC-2022-0073-DRAFT-0013-5 requirements defined and stated in ASME/ANS RA-S-1.4-2021, as e ndorsed in RG 1.247 (for trial use) Commented [A57]: NRC-2022-0073-DRAFT-0013-1 with clarifications and qualifications, provide an acceptable a pproach for developing PRA logic models. NRC-2022-0073-DRAFT-0013-2 NRC-2022-0073-DRAFT-0013-5 A.3.5 Regarding PRA acceptability, Section 4.3.11 of ASME/ANS R A-S-1.4-2021, which is endorsed NRC-2022-0073-DRAFT-0013-7 in RG 1.247 (for trial use) with exceptions, provides an accept able approach for performing a hazards NRC-2022-0073-DRAFT-0013-10 NRC-2022-0073-DRAFT-0013-11 screening analysis. Commented [A58]: NRC-2022-0073-DRAFT-0013-6

A.3.6 Regarding PRA acceptability, risk-informed supplemental e valuations may be used to disposition Commented [A59]: NRC-2022-0073-DRAFT-0013-1 certain radiological sources, hazards, or POSs. NUREG-1855, Rev ision 1, Guidance on the Treatment of NRC-2022-0073-DRAFT-0013-2 NRC-2022-0073-DRAFT-0013-5 Uncertainties Associated with PRAs in Risk-Informed Decisionmaking (Ref. A-19), provides a generally NRC-2022-0073-DRAFT-0013-7 acceptable approach for developing risk-informed supplemental e valuations. Section 1.3 of NUREG-1855 NRC-2022-0073-DRAFT-0013-10 notes that the process described in NUREG-1855 is applicable to non-LWRs and reactors in the NRC-2022-0073-DRAFT-0013-11 design stage; however, the screening criteria and the specific sources of uncertainty may not be Commented [A60]: NRC-2022-0073-DRAFT-0013-1 applicable. Consequently, non-LWR CP applicants who use the gu idance in NUREG-1855 to develop NRC-2022-0073-DRAFT-0013-2 risk-informed supplemental evaluations should (1) describe and justify the use of reactor-technology-NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 specific screening criteria, and (2) explain how specific sourc es of uncertainty were identified and NRC-2022-0073-DRAFT-0013-10 addressed. NRC-2022-0073-DRAFT-0013-11

A.3.7 The CP applicant may disposition certain hazards by credi ting DBHLs in lieu of explicitly modeling these hazards in the PRA or accounting for them throug h a risk-informed supplementary evaluation. NEI 1804, Revision 1, Section 3.2.2, Task 6, p. 14, states:

In many cases, it is expected that the initial selection of SR SSCs [safety-related structures, systems, and components] and selection of the DBAs [design-basis accidents]

will be based on a PRA that includes internal events but has no t yet been expanded to address external hazards. With the understanding that SR SSCs a re required to be capable of performing their RSFs [required safety functions] in respons e to external events within

4 The non-LWR PRA standard uses the terms requirement, require, and other similar mandatory language. However, the use of this language in this RG does not imply that this RG imposes any regulatory requirement or suggest that these standards are the only way to meet the statutory and regulatory requirements.

RG 1.253, App. A, Page A-5 the DBEHL [design-basis external hazard levels], there will be no new DBAs introduced by external hazards.

NEI 2107, Revision 1, Section 6.1.1 clarified NEI 1804 by def ining and using the term DBHL rather than DBEHL. Specifically, DBHLs address both traditional extern al hazards (e.g., seismic events, external floods, high winds) and internal hazards (e.g., intern al fires, internal floods, turbine missiles, and high energy line breaks). Consistent with Section C.1 in RG 1.2 33, the scope of the PRA, when completed, should cover a full set of internal and external ini tiating events and provide an estimate of radiological consequences when the design is completed and site characteristics are defined. Commented [A61]: NRC-2022-0073-DRAFT-0013-1 NRC-2022-0073-DRAFT-0013-2 A.4 PRA Elements NRC-2022-0073-DRAFT-0013-4 NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 A.4.1 Regarding PRA acceptability, Table A-1 shows which PRA el ements defined in Staff Position NRC-2022-0073-DRAFT-0013-10 C.1.3 of RG 1.247 apply to the minimally acceptable PRA scope a nd additional PRA elements that may NRC-2022-0073-DRAFT-0013-11 be used to fully implement the LMP methodology at the CP stage. Commented [A62]: NRC-2022-0073-DRAFT-0013-1 NRC-2022-0073-DRAFT-0013-2 Table A-1. PRA Elements for Non-LWR CP Applications Based on th e LMP Methodology NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 Minimally Acceptable PRA Additional PRA Elements NRC-2022-0073-DRAFT-0013-10 Identifier PRA Element Identifier PRA Element NRC-2022-0073-DRAFT-0013-11 C.1.3.2 (IE) Initiating Event Analysis C.1.3.1 (POS) Plant Operating State Analysis C.1.3.3 (ES) Event Sequence Analysis C.1.3.8 (IF) Internal Flood PRA C.1.3.4 (SC) Success Criteria Development C.1.3.9 (F) Internal Fire PRA C.1.3.5 (SY) Systems Analysis C.1.3.10 (S) Seismic PRA C.1.3.6 (HR) Human Reliability A n a lysis C.1.3.12 (W) High Wind PRA C.1.3.7 (DA) Data Analysis C.1.3.13 (XF) External Flooding PRA C.1.3.11 (HS) Hazard Screening Analysis C.1.3.14 (O) Other Haza rds PRA C.1.3.15 (ESQ) Event Sequence Quantification C.1.3.16 (MS) Mechanistic Source Term Analysis C.1.3.17 (RC) Radiological Consequence Analysis C.1.3.18 (RI) Risk Integration

A.4.2 Consistent with Section C.2.1.1 of NEI 21-07, Revision 1, the PRA elements should be developed Commented [A63]: NRC-2022-0073-DRAFT-0013-1 to conform with the high-level requirements and associated supp orting requirements of ASME/ANS NRC-2022-0073-DRAFT-0013-2 RA-S-1.4-2021, as endorsed with clarifications and qualificatio ns in Appendix A of RG 1.247. Consistent NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 with Staff Position C.2.1 in RG 1.247, all high-level requireme nts for a given PRA element should be NRC-2022-0073-DRAFT-0013-10 met. NRC-2022-0073-DRAFT-0013-11 Commented [A64]: NRC-2022-0073-DRAFT-0013-6 A.5 PRA Level of Detail

A.5.1 Consistent with Section C.2.1.1 of NEI 21-07, Revision 1, the CP PRA level of detail should be Commented [A65]: NRC-2022-0073-DRAFT-0013-1 commensurate with the preliminary plant design and site charact eristics described in the PSAR. NRC-2022-0073-DRAFT-0013-2 NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 A.5.2 The level of detail in a CP PRA should be established usi ng the process provided in Section 3 of NRC-2022-0073-DRAFT-0013-10 ASME/ANS RA-S-1.4-2021, Risk Assessment Application Process. If an applicant meets the NRC-2022-0073-DRAFT-0013-11 provisions of Table A-2, the NRC staff would consider this to r esult in an acceptable level of detail for the PRA logic model and hazard screening analyses supporting an LMP -based CP application. To the extent the provisions in Table A-2 cannot be met due to the maturity o f preliminary plant design and information about site characteristics, the application should justify the adequacy of the internal events PRA logic model for the reactor in the at-power POSs to support the imple mentation of the LMP methodology. If the maturity of the preliminary plant design and information about site characteristics is sufficient to support

RG 1.253, App. A, Page A-6 the development of a PRA logic model that includes other intern al and external hazards or other POSs, the applicant should consider applying the provisions of Table A-3. Commented [A66]: NRC-2022-0073-DRAFT-0013-9 NRC-2022-0073-DRAFT-0007-4 A.6 Plant Representation and PRA Configuration Control NRC-2022-0073-DRAFT-0007-5 Commented [A67]: NRC-2022-0073-DRAFT-0013-9 A.6.1 Consistent with Section C.2.1.1 of NEI 21-07, Revision 1, the CP applicant should establish a NRC-2022-0073-DRAFT-0007-4 PRA configuration control program to ensure that the CP PRA rea sonably represents the preliminary plant NRC-2022-0073-DRAFT-0007-5 design and site characteristics described in the PSAR. Commented [A68]: NRC-2022-0073-DRAFT-0013-1 NRC-2022-0073-DRAFT-0013-2 NRC-2022-0073-DRAFT-0013-5 A.6.2 Regarding PRA acceptability, Section 5 of ASME/ANS RA-S-1.4-2021, which is endorsed in NRC-2022-0073-DRAFT-0013-7 RG 1.247 (for trial use) with exceptions, provides one acceptab le approach for establishing a PRA NRC-2022-0073-DRAFT-0013-10 configuration control program. NRC-2022-0073-DRAFT-0013-11 Commented [A69]: NRC-2022-0073-DRAFT-0013-1 A.6.3 Regarding PRA acceptability, consistent with the discussi on provided in NUREG-0800, NRC-2022-0073-DRAFT-0013-2 NRC-2022-0073-DRAFT-0013-5 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR NRC-2022-0073-DRAFT-0013-7 Edition (Ref. A-20), Section 19.0, Item 9, page 19.0-12, the P RA configuration control program may be NRC-2022-0073-DRAFT-0013-10 a stand-alone program or included within the quality assurance program required by 10 CFR 50.34(a)(7). NRC-2022-0073-DRAFT-0013-11 Commented [A70]: NRC-2022-0073-DRAFT-0013-1 A.7 PRA Documentation NRC-2022-0073-DRAFT-0013-2 NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 A.7.1 NEI 2107, Revision 1, as endorsed with additions and cla rifications in the main body of this RG, NRC-2022-0073-DRAFT-0013-10 provides an acceptable approach and format for providing CP PRA information in the SAR. Consistent NRC-2022-0073-DRAFT-0013-11 with staff position C.4.2 of RG 1.247 on PRA documentation for an application, the PSAR should provide a summary justification of the acceptability of the CP PRA. Thi s justification should summarize why, commensurate with the level of maturity of the facility design, the scope, level of detail, elements of the PRA, and plant representation described in the CP PRA is suffic ient to implement the LMP methodology for the CP application. To this end, the summary justification should describe, but is not limited to, the following: Commented [A71]: NRC-2022-0073-DRAFT-0013-1 NRC-2022-0073-DRAFT-0013-11

• How the CP PRA was developed in accordance with the guidance pr ovided in this appendix and the portions of RG 1.247 and the ASME/ANS non-LWR PRA consensus sta ndard (ASME/ANS RA-S-1.4-2021) the applicant has followed in developing the CP PRA. Commented [A72]: I've gone back to "Robert Weisman 2d" as the name for these comments because Word, in its infinite

• How a PRA configuration control program has been used to ensure the CP PRA represents the as-wisdom, has assigned "Robert Weisman penultimate" a color designed, as-to-be-built, as-to-be-operated facility described in the CP application. essentially indistinguishable from the color assigned to Anders's comments.

• How the PRA configuration control program will ensure that the PRA (i.e., PRA logic model, Commented [A73R72]: Agreed with edits.

screening analyses, and risk-informed supplemental evaluations) supporting the OL application will represent the as-built, as-to-be-operated facility; account for all radiological sources, all hazards, and all plant operating states; and meet all applicable staff posit ions in RG 1.247 and technical elements in ASME/ANS RA-S-1.4-2021 that the CP PRA did not meet.

• How the applicants self-assessment of the CP PRA was performed consistent with the related staff positions in Section A.8 of this regulatory guide.

A.7.2 Regarding acceptability of archival documentation confirm ing the acceptability of the CP PRA, Commented [A74]: NRC-2022-0073-DRAFT-0013-1 Staff Position C.4.1 in RG 1.247 provides an acceptable approac h for developing and preserving PRA NRC-2022-0073-DRAFT-0013-2 archival information. PRA documentation providing the detailed justification for the acceptability of the NRC-2022-0073-DRAFT-0013-5 PRA should be maintained in archival PRA documentation and, at a minimum, should include the items NRC-2022-0073-DRAFT-0013-7 NRC-2022-0073-DRAFT-0013-10 described in staff position C.4.1 of RG 1.247. NRC-2022-0073-DRAFT-0013-11

RG 1.253, App. A, Page A-7 A.7.3 Consistent with the discussion provided in NUREG-0800, Se ction 19.0, Item 9, page 19.0-12, PRA archival information may be controlled by a stand-alone pro gram, or the quality assurance program required by 10 CFR 50.34(a)(7).

A.8 Demonstrating PRA Acceptability

A.8.1 The guidance in DANU-ISG-2022-05, Organization and Human -System Considerations (Ref. A-21), Section 11.1.1, provides an acceptable approach fo r describing key management responsibilities for developing the PRA.

A.8.2 The guidance in DANU-ISG-2022-05, Section 11.1.1.1, provi des an acceptable approach for describing the ability of the CP applicants technical staff to develop the PRA.

A.8.3 Regarding PRA acceptability, the CP applicant should cond uct a self-assessment to demonstrate Commented [A75]: NRC-2022-0073-DRAFT-0013-12 that all PRA logic models, screening analyses, and risk-informe d supplemental analyses have been Commented [A76]: NRC-2022-0073-DRAFT-0013-1 developed and used in a technically acceptable manner, includin g the appropriateness of assumptions and NRC-2022-0073-DRAFT-0013-2 approximations. The self-assessment, thus, should provide a bas is for asserting that the CP PRA is NRC-2022-0073-DRAFT-0013-5 acceptable for implementing the LMP methodology leading up to s ubmittal of the CP application. To this NRC-2022-0073-DRAFT-0013-7 end, the self-assessment should review: NRC-2022-0073-DRAFT-0013-10 NRC-2022-0073-DRAFT-0013-11

1. The comprehensive and systematic search used to identify rad iological sources, POSs, and hazards.

2. The PRA logic models (including the scope, level of detail, and elements), screening analyses, risk-informed supplemental evaluations, and credit for DBHLs.

3. The CP applicants PRA configuration control program used to ensure that CP PRA logic models, screening analyses, and risk-informed supplemental analyses rep resent the as-designed, as-to-be-built, and as-to-be-operated plant.

A.8.4 Regarding PRA acceptability, the guidance in NEI 2009, R evision 1, Performance of PRA Peer Commented [A77]: NRC-2022-0073-DRAFT-0013-1 Reviews Using the ASME/ANS Advanced Non-LWR PRA Standard, (Ref. A-22), Sections 3.2, A.3.1, NRC-2022-0073-DRAFT-0013-2 and A.3.2, which is endorsed in RG 1.247 with no exceptions, pr ovides an acceptable approach for NRC-2022-0073-DRAFT-0013-5 NRC-2022-0073-DRAFT-0013-7 performing a self-assessment. NRC-2022-0073-DRAFT-0013-10 NRC-2022-0073-DRAFT-0013-11 A.8.5 Regarding PRA acceptability, in addition to a self-assess ment, a CP applicant may have the Commented [A78]: NRC-2022-0073-DRAFT-0013-1 comprehensive and systematic search, some or all PRA elements, some or all screening analyses, or the NRC-2022-0073-DRAFT-0013-2 PRA configuration control program peer reviewed prior to submit tal of the CP application.5 Section 6 of NRC-2022-0073-DRAFT-0013-5 ASME/ANS RA-S-1.4-2021, which is endorsed in Staff Position C.2.2 in RG 1.247 with exceptions, and NRC-2022-0073-DRAFT-0013-7 NRC-2022-0073-DRAFT-0013-10 NEI 2009, Revision 1, which is endorsed in RG 1.247 without ex ception, provide an acceptable approach NRC-2022-0073-DRAFT-0013-11 for performing a peer review. Commented [A79]: NRC-2022-0073-DRAFT-0013-6

The purpose of a peer review is to determine whether the releva nt high-level requirements and associated supporting requirements established in ASME/ANS RA-S-1.4-2021, as endorsed in RG 1.247 with Commented [A80]: NRC-2022-0073-DRAFT-0013-6 exceptions, have been met. The peer review should also confirm that the technical aspects of the CP PRA have been developed in a technically correct manner and assess the appropriateness of assumptions and

5 High-level requirements and supporting requirements related to the comprehensive and systematic search for radionuclide sources, POSs, and hazards appear in various locations throughout ASME/ANS RA-S-1.4-2021. However, the standard does not provide high-level requirements and supporting requirements for risk-informed supplemental evaluations.

RG 1.253, App. A, Page A-8 approximations used in the CP PRA. As a result, completion of a peer review may reduce the need for an in-depth staff review of the CP PRA.

RG 1.253, App. A, Page A-9 Appendix A Tables A-2 and A-3 Notes

The staff applied the risk assessment application process provi ded in Section 3 of ASME/ANS RA-S-1.4-2021 to determine the applicability of supporting requ irements, as defined in the ASME/ANS RA-S-1.4-2021, for a non-LWR CP application that implements the LMP methodology. The staffs application of this process considered a wide range of maturiti es of design information that may be submitted to the NRC in a CP application. The results, which re flect the provisions of ASME/ANS RA-S-1.4-2021 that an applicant should follow for a fully mature design, are presented in Tables A-2 and A-3, which follow immediately below. The CP applicant should us e Tables A-2 and A-3 as guidance to demonstrate the acceptability of the scope and level of detail of the PRA logic model, consistent with the maturity of the design. Alternatively, the CP applicant may per form a separate analysis using the process provided in Section 3 of ASME/ANS RA-S-1.4-2021 and justify any departures from or alternatives to Tables A-2 and A-3, depending on the maturity of design informa tion for a given applicant. As such, an applicant need not use Tables A-2 and A-3 if an applicant justi fies a different combination of applicable ASME/ANS supporting requirements for its application.

As referenced in staff position A.5.2, Table A-2 shows the alph anumeric identifiers of the underlined high-level requirements and their related supporting requiremen ts in tabular format from ASME/ANS RA-S-1.4-2021 for a minimally acceptable PRA logic model and ha zard screening analysis for a CP application under 10 CFR Part 50. As discussed in position A.3. 3, the minimally acceptable scope of the CP PRA logic model should include the internal events hazard fo r the reactor in the at-power POS. Table A-3 shows the related alphanumeric identifiers for a PRA logic model associated with a design that includes other internal and external hazards or other POSs for a CP application. The in-line entry associated with each ASME/ANS supporting requirement is either Yes, CC-I, or CC-II. A Yes entry indicates the supporting requirements for CC-I and CC-II (i.e., the two types of capability categories in ASME/ANS RA-S-1.4-2021) are applicable because the supportin g requirement is the same for both capability categories. An entry of either CC-I or CC-II ind icates the minimum applicable supporting requirement associated with that capability category. The staff did not include inapplicable supporting Commented [A81]: NRC-2022-0073-DRAFT-0013-9 requirements in these tables. NRC-2022-0073-DRAFT-0007-4 NRC-2022-0073-DRAFT-0007-5

RG 1.253, App. A, Page A-10 Table A-2. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to PRA Elements Used to Develop a Minimally Accept able PRA (1 of 4)

C.1.3.2 (IE) HLR-IE-D ES-C9 CC-II SY-A4 Yes HLR-IE-A IE-D1 Yes ES-C10 Yes SY-A6 CC-II IE-A1 Yes IE-D2 Yes ES-C11 Yes SY-A7 CC-I IE-A2 Yes IE-D3 Yes SY-A8 Yes IE-A4 Yes HLR-ES-D SY-A9 Yes IE-A5 Yes C.1.3.3 (ES) E S - D 1 Ye s S Y-A 1 1 Ye s I E - A 6 Ye s H L R - E S - A E S - D 2 Ye s S Y-A 1 2 Ye s IE-A8 CC-II ES-A1 Yes ES-D3 Yes SY-A13 Yes IE-A9 CC-II ES-A2 Yes SY-A14 Yes IE-A10 CC-I ES-A3 Yes C.1.3.4 (SC) SY-A15 Yes IE-A11 Yes ES-A4 Yes HLR-SC-A SY-A16 Yes IE-A12 CC-I ES-A5 Yes SC-A1 Yes SY-A17 Yes IE-A14 Yes ES-A6 Yes SC-A2 Yes SY-A18 Yes IE-A15 Yes ES-A7 Yes SC-A3 CC-II SY-A19 Yes I E - A 1 6 Ye s E S - A 8 Ye s S C - A 4 Ye s S Y-A 2 0 Ye s I E - A 1 7 Ye s E S - A 9 Ye s S C - A 5 Ye s S Y-A 2 1 Ye s IE-A18 Yes ES-A10 CC-II SC-A6 Yes SY-A22 Yes ES-A11 Yes SC-A7 CC-II SY-A23 Yes HLR-IE-B ES-A12 CC-II SC-A8 Yes SY-A24 Yes IE-B1 Yes ES-A13 Yes SC-A9 Yes SY-A25 Yes IE-B2 Yes ES-A14 Yes SC-A10 Yes SY-A26 Yes I E - B 3 Ye s E S - A 1 5 Ye s S C - A 1 1 Ye s S Y-A 2 7 Ye s IE-B4 CC-II SY-A28 Yes IE-B5 Yes HLR-ES-B HLR-SC-B SY-A29 CC-I IE-B6 Yes ES-B1 Yes SC-B1 CC-II SY-A30 Yes ES-B2 Yes SC-B2 Yes SY-A31 Yes HLR-IE-C ES-B3 Yes SC-B3 Yes SY-A32 Yes I E - C 2 Ye s E S - B 4 Ye s S C - B 4 Ye s S Y-A 3 3 Ye s I E - C 4 Ye s E S - B 5 Ye s S C - B 5 Ye s I E - C 5 Ye s E S - B 6 Ye s S C - B 6 Ye s H L R - S Y-B IE-C7 Yes ES-B7 Yes SC-B7 CC-I SY-B1 CC-I IE-C8 Yes ES-B8 Yes SC-B8 Yes SY-B2 CC-I I E - C 9 Ye s E S - B 9 Ye s S C - B 9 Ye s S Y-B 3 Ye s IE-C10 CC-II ES-B10 Yes SC-B10 Yes SY-B4 Yes IE-C11 Yes SY-B5 Yes IE-C12 Yes HLR-ES-C HLR-SC-C SY-B6 Yes IE-C13 Yes ES-C1 Yes SC-C1 Yes SY-B7 CC-I IE-C14 Yes ES-C2 Yes SC-C2 Yes SY-B8 Yes IE-C15 Yes ES-C3 Yes SC-C3 Yes SY-B9 Yes IE-C16 Yes ES-C4 Yes SY-B10 Yes IE-C17 CC-II ES-C5 Yes C.1.3.5 (SY) SY-B11 CC-II IE-C18 Yes ES-C6 Yes HLR-SY-A SY-B12 Yes IE-C19 CC-II ES-C7 CC-I SY-A1 Yes SY-B13 Yes Commented [A82]: NRC-2022-0073-DRAFT-0007-1 ES-C8 Yes SY-A2 Yes SY-B14 Yes NRC-2022-0073-DRAFT-0007-2 S Y-A 3 Ye s S Y-B 1 5 Ye s NRC-2022-0073-DRAFT-0007-3

RG 1.253, App. A, Page A-11 Table A-2. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to PRA Elements Used to Develop a Minimally Accept able PRA (2 of 4)

SY-B16 Ye s H L R - H R - E C.1.3.7 (DA) HS-A3 Yes SY-B17 Yes HR-E2 Yes HLR-DA-A HS-A4 Yes HR-E3 Yes DA-A1 Yes HLR-SY-C HR-E4 Yes DA-A2 Yes HLR-HS-B SY-C1 Yes HR-E6 CC-I DA-A3 Yes HS-B1 Yes SY-C2 Yes HR-E7 CC-I DA-A4 Yes HS-B2 Yes SY-C3 Yes HR-E8 Yes DA-A5 Yes HS-B4 Yes HR-E9 Yes DA-A6 Yes HS-B5 Yes C.1.3.6 (HR) HS-B6 Yes HLR-HR-A HLR-HR-F HLR-DA-B HS-B7 Yes HR-A1 Yes HR-F1 Yes DA-B1 CC-II HR-A2 Yes HR-F2 Yes DA-B2 Yes HLR-HS-C HR-A3 Yes HR-F3 CC-II HS-C1 Yes HR-A4 Yes HR-F4 CC-II HLR-DA-C HS-C2 Yes HR-A5 Yes HR-F5 Yes DA-C1 Yes HS-C3 Yes HR-A6 Yes DA-C2 Yes HS-C4 Yes HR-A7 Yes HLR-HR-G DA-C9 CC-I HS-C5 Yes Commented [A83]: NRC-2022-0073-DRAFT-0007-1 HR-A8 Yes HR-G1 CC-I DA-C14 Yes HS-C6 Yes NRC-2022-0073-DRAFT-0007-2 HR-A9 Yes HR-G2 Yes DA-C15 Yes HS-C7 Yes NRC-2022-0073-DRAFT-0007-3 HR-A10 Yes HR-G3 Yes DA-C17 CC-I HS-C8 Yes HR-G4 CC-I DA-C19 Yes HS-C9 Yes HLR-HR-B HR-G5 CC-II DA-C20 Yes HS-C10 Yes HR-B1 Yes HR-G6 CC-II DA-C21 Yes HS-C11 Yes H R - B 2 Ye s H R - G 7 Ye s D A - C 2 3 Ye s H S - C 1 2 Ye s HR-B3 Yes HR-G8 CC-I D A - C 2 5 Ye s H S - C 1 3 Ye s Commented [A84]: NRC-2022-0073-DRAFT-0007-1 HR-G10 Yes HS-C14 Yes NRC-2022-0073-DRAFT-0007-2 HLR-HR-C HR-G11 Yes HLR-DA-D NRC-2022-0073-DRAFT-0007-3 HR-C1 Yes HR-G12 Yes DA-D1 CC-I HLR-HS-D Commented [A85]: NRC-2022-0073-DRAFT-0007-1 HR-C2 Yes HR-G13 Yes DA-D2 Yes HS-D1 Yes NRC-2022-0073-DRAFT-0007-2 HR-C3 Yes HR-G14 CC-II DA-D3 CC-II NRC-2022-0073-DRAFT-0007-3 HR-C4 CC-II HR-G15 Yes DA-D5 Yes HLR-HS-E HR-C5 Yes HR-G16 Yes DA-D6 Yes HS-E1 Yes HR-C6 Yes DA-D7 CC-II HS-E2 Yes HLR-HR-H DA-D8 CC-I HS-E3 Yes HLR-HR-D HR-H1 Yes DA-D9 Yes HR-D1 Yes HR-H2 Yes C.1.3.15 (ESQ)

HR-D2 CC-I HR-H3 Yes HLR-DA-E HLR-ESQ-A HR-D3 Yes HR-H4 CC-I DA-E1 Yes ESQ-A1 Yes HR-D4 CC-I HR-H5 Yes DA-E2 Yes ESQ-A2 Yes HR-D5 Yes HR-H6 Yes DA-E3 Yes ESQ-A3 Yes HR-D7 Yes ESQ-A4 Yes HR-D8 CC-II HLR-HR-I C.1.3.11 (HS) ESQ-A5 CC-II HR-D9 Yes HR-I1 Yes HLR-HS-A ESQ-A6 Yes H R - D 1 0 Ye s H R - I 2 Ye s H S - A 1 Ye s E S Q - A 7 Ye s HR-I3 Yes HS-A2 Yes

RG 1.253, App. A, Page A-12 Table A-2. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to PRA Elements Used to Develop a Minimally Accept able PRA (3 of 4)

ESQ-A8 CC-I HLR-ESQ-E HLR-MS-E HLR-RCPA-C ESQ-A9 CC-I ESQ-E1 Yes MS-E1 Yes RCPA-C1 Yes ESQ-E2 CC-II MS-E2 Yes RCPA-C2 Yes HLR-ESQ-B MS-E3 Yes ESQ-B1 Yes HLR-ESQ-F MS-E4 Yes HLR-RCME-A ESQ-B2 Yes ESQ-F1 Yes RCME-A1 Yes ESQ-B3 Yes ESQ-F2 Yes C.1.3.17 (RC) RCME-A2 CC-I Commented [A86]: NRC-2022-0073-DRAFT-0007-1 ESQ-B4 Yes ESQ-F3 Yes HLR-RCRE-A RCME-A3 CC-II NRC-2022-0073-DRAFT-0007-2 ESQ-B5 Yes ESQ-F4 Yes RCRE-A1 Yes RCME-A4 CC-II NRC-2022-0073-DRAFT-0007-3 ESQ-B6 Yes ESQ-F5 Yes RCRE-A2 Yes RCME-A5 CC-II ESQ-B7 Yes RCRE-A3 Yes RCME-A6 CC-II ESQ-B8 Yes C.1.3.16 (MS) RCME-A7 CC-II ESQ-B9 Yes HLR-MS-A HLR-RCRE-B RCME-A8 Yes ESQ-B10 Yes MS-A1 Yes RCRE-B1 Yes RCME-A9 Yes MS-A2 Yes RCRE-B2 Yes RCME-A10 Yes HLR-ESQ-C MS-A3 CC-II ESQ-C1 Yes MS-A4 Yes HLR-RCRE-C HLR-RCME-B ESQ-C2 Yes MS-A5 Yes RCRE-C1 Yes RCME-B1 Yes ESQ-C3 Yes RCME-B2 Yes ESQ-C4 Yes HLR-MS-B HLR-RCPA-A ESQ-C5 CC-II MS-B1 CC-I RCPA-A1 Yes HLR-RCAD-A ESQ-C6 CC-II MS-B2 Yes RCPA-A2 Yes RCAD-A1 CC-II ESQ-C7 CC-I MS-B3 Yes RCPA-A3 Yes RCAD-A2 CC-II ESQ-C8 CC-I MS-B4 Yes RCPA-A4 CC-II RCAD-A3 CC-II Commented [A87]: NRC-2022-0073-DRAFT-0007-1 ESQ-C9 CC-I MS-B5 Yes RCPA-A5 Yes RCAD-A4 Yes NRC-2022-0073-DRAFT-0007-2 ESQ-C10 Yes MS-B6 Yes RCPA-A6 Yes RCAD-A5 Yes NRC-2022-0073-DRAFT-0007-3 ESQ-C11 Yes MS-B7 Yes RCPA-A7 Yes RCAD-A6 Yes ESQ-C12 Yes RCPA-A8 Yes RCAD-A7 Yes ESQ-C13 Yes HLR-MS-C RCPA-A9 Yes ESQ-C14 CC-I MS-C1 CC-I RCPA-A10 Yes HLR-RCAD-B ESQ-C15 CC-II MS-C2 CC-I RCPA-A11 Yes RCAD-B1 Yes ESQ-C16 Yes MS-C3 CC-I RCPA-A12 Yes RCAD-B2 CC-I ESQ-C17 Yes MS-C4 CC-II RCPA-A13 Yes MS-C5 Yes HLR-RCAD-C HLR-ESQ-D MS-C6 Yes HLR-RCPA-B RCAD-C1 Yes ESQ-D1 Yes MS-C7 Yes RCPA-B1 CC-II RCAD-C2 CC-II ESQ-D2 Yes RCPA-B2 CC-II RCAD-C3 Yes ESQ-D3 Yes HLR-MS-D RCPA-B3 CC-II RCAD-C4 Yes ESQ-D4 CC-II MS-D1 Yes RCPA-B4 Yes RCAD-C5 Yes ESQ-D5 Yes MS-D2 CC-II RCPA-B6 Yes ESQ-D6 CC-II MS-D3 Yes RCPA-B7 Yes HLR-RCAD-D ESQ-D7 Yes MS-D4 CC-II RCAD-D1 CC-II ESQ-D8 Yes RCAD-D2 CC-II RCAD-D3 Yes

RG 1.253, App. A, Page A-13 Table A-2. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to PRA Elements Used to Develop a Minimally Accept able PRA (4 of 4)

HLR-RCAD-E HLR-RCHE-C HLR-RI-D RCAD-E1 CC-II RCHE-C1 Yes RI-D1 Yes RCAD-E2 CC-II RCHE-C2 Yes RI-D2 Yes RCAD-E3 CC-II RCAD-E4 CC-II HLR-RCQ-A RCAD-E5 Yes RCQ-A1 Yes RCAD-E6 Yes RCQ-A2 Yes RCQ-A3 Yes HLR-RCAD-F RCAD-F1 Yes HLR-RCQ-B RCAD-F2 Yes RCQ-B1 Yes RCQ-B2 Yes HLR-RCDO-A RCQ-B3 Ye s RCDO-A1 Yes RCDO-A2 Yes HLR-RCQ-C RCDO-A3 Yes RCQ-C1 Yes RCDO-A4 CC-II RCQ-C2 CC-I Commented [A88]: NRC-2022-0073-DRAFT-0007-1 RCDO-A5 Yes NRC-2022-0073-DRAFT-0007-2 RCDO-A6 Yes HLR-RCQ-D NRC-2022-0073-DRAFT-0007-3 RCDO-A7 CC-II RCQ-D1 Yes RCDO-A8 Yes RCQ-D2 Yes RCDO-A9 Yes RCQ-D3 Yes RCDO-A10 Yes C.1.3.18 (RI)

HLR-RCDO-B HLR-RI-A RCDO-B1 CC-II RI-A1 Yes RCDO-B2 Yes RI-A2 Yes RI-A3 Yes HLR-RCDO-C RI-A4 Yes RCDO-C1 Yes RI-A5 Yes RCDO-C2 Yes HLR-RI-B HLR-RCHE-A RI-B1 Yes RCHE-A1 Yes RI-B2 CC-II RCHE-A2 CC-I RI-B3 CC-II Commented [A89]: NRC-2022-0073-DRAFT-0007-1 RCHE-A3 CC-I RI-B4 Yes NRC-2022-0073-DRAFT-0007-2 RCHE-A4 Yes RI-B5 Yes NRC-2022-0073-DRAFT-0007-3 RCHE-A5 Yes RI-B6 Yes Commented [A90]: NRC-2022-0073-DRAFT-0007-1 RCHE-A6 Yes RI-B7 Yes NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 HLR-RCHE-B HLR-RI-C RCHE-B1 Yes RI-C1 Yes RCHE-B2 Yes RI-C2 Yes RCHE-B3 Yes RI-C3 Yes RI-C4 CC-II

RG 1.253, App. A, Page A-14 Table A-3. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to Additional PRA Elements (1 of 6)

C.1.3.1 (POS) FLPP-B7 Yes FLSN-A20 Yes HLR-FLPR-C HLR-POS-A FLPP-B8 Yes FLSN-A21 Yes FLPR-C1 Yes POS-A1 CC-II FLPR-C2 Yes POS-A2 Yes HLR-FLPP-C HLR-FLSN-B FLPR-C3 Yes POS-A3 Yes FLPP-C1 Yes FLSN-B1 Yes POS-A5 Yes FLPP-C2 Yes FLSN-B2 Yes HLR-FLHR-A POS-A8 Yes FLPP-C3 Yes FLSN-B3 Yes FLHR-A1 CC-I Commented [A91]: NRC-2022-0073-DRAFT-0007-1 POS-A9 Yes FLHR-A2 CC-I NRC-2022-0073-DRAFT-0007-2 POS-A10 Yes HLR-FLSO-A HLR-FLEV-A NRC-2022-0073-DRAFT-0007-3 POS-A11 Yes FLSO-A1 Yes FLEV-A1 CC-II HLR-FLHR-B Commented [A92]: NRC-2022-0073-DRAFT-0007-1 POS-A12 Yes FLSO-A2 Yes FLEV-A2 CC-I FLHR-B1 Yes NRC-2022-0073-DRAFT-0007-2 POS-A13 Yes FLSO-A3 Yes FLEV-A3 Yes FLHR-B2 CC-II NRC-2022-0073-DRAFT-0007-3 FLSO-A4 Yes FLEV-A4 Yes FLHR-B3 CC-II HLR-POS-B FLSO-A5 Yes POS-B1 CC-II FLSO-A6 Yes HLR-FLEV-B HLR-FLHR-C POS-B2 Yes FLSO-A7 Yes FLEV-B1 CC-I FLHR-C1 CC-II Commented [A93]: NRC-2022-0073-DRAFT-0007-1 POS-B3 Yes FLSO-A8 Yes FLEV-B2 Yes NRC-2022-0073-DRAFT-0007-2 POS-B4 Yes FLSO-A9 Yes FLEV-B3 CC-I HLR-FLHR-D NRC-2022-0073-DRAFT-0007-3 POS-B5 CC-I FLEV-B4 CC-I FLHR-D1 CC-I POS-B6 Yes HLR-FLSO-B FLEV-B5 Yes FLHR-D2 Yes POS-B7 Yes FLSO-B1 Yes FLEV-B6 Yes FLHR-D3 Yes POS-B8 Yes FLSO-B2 Yes FLEV-B7 Yes FLSO-B3 Yes HLR-FLHR-E HLR-POS-C HLR-FLEV-C FLHR-E1 Yes POS-C1 Yes HLR-FLSN-A FLEV-C1 Yes FLHR-E2 Yes POS-C2 Yes FLSN-A1 Yes FLEV-C2 Yes FLHR-E3 Yes POS-C3 Yes FLSN-A2 Yes FLEV-C3 Yes POS-C4 Yes FLSN-A3 Yes HLR-FLESQ-A FLSN-A4 Yes HLR-FLPR-A FLESQ-A1 Yes HLR-POS-D FLSN-A5 Yes FLPR-A1 Yes FLESQ-A2 Yes POS-D1 Yes FLSN-A6 CC-II FLPR-A2 Yes FLESQ-A3 CC-I Commented [A94]: NRC-2022-0073-DRAFT-0007-1 POS-D2 Yes FLSN-A7 Yes FLPR-A3 Yes FLESQ-A4 CC-I NRC-2022-0073-DRAFT-0007-2 POS-D3 Yes FLSN-A8 CC-II FLESQ-A5 Yes NRC-2022-0073-DRAFT-0007-3 FLSN-A9 CC-I HLR-FLPR-B FLESQ-A6 Yes Commented [A95]: NRC-2022-0073-DRAFT-0007-1 C.1.3.8 (FL) FLSN-A10 CC-I FLPR-B1 Yes FLESQ-A7 Yes NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 HLR-FLPP-A FLSN-A11 Yes FLPR-B2 Yes FLESQ-A8 Yes Commented [A96]: NRC-2022-0073-DRAFT-0007-1 FLPP-A1 Yes FLSN-A12 Yes FLPR-B3 CC-II NRC-2022-0073-DRAFT-0007-2 FLSN-A13 Yes FLPR-B4 CC-II HLR-FLESQ-B NRC-2022-0073-DRAFT-0007-3 HLR-FLPP-B FLSN-A14 Yes FLPR-B5 CC-II FLESQ-B1 Yes Commented [A97]: NRC-2022-0073-DRAFT-0007-1 FLPP-B1 Yes FLSN-A15 Yes FLPR-B6 CC-I NRC-2022-0073-DRAFT-0007-2 FLPP-B2 Yes FLSN-A16 Yes FLPR-B7 CC-I HLR-FLESQ-C NRC-2022-0073-DRAFT-0007-3 FLPP-B4 Yes FLSN-A17 Yes FLPR-B8 CC-II FLESQ-C1 Yes Commented [A98]: NRC-2022-0073-DRAFT-0007-1 FLPP-B5 Yes FLSN-A18 Yes FLPR-B9 Yes NRC-2022-0073-DRAFT-0007-2 FLPP-B6 Yes FLSN-A19 Yes FLPR-B10 Yes NRC-2022-0073-DRAFT-0007-3 Commented [A99]: NRC-2022-0073-DRAFT-0007-1 NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3

RG 1.253, App. A, Page A-15 Table A-3. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to Additional PRA Elements (2 of 6)

HLR-FLESQ-D FES-B2 CC-II HLR-FPRM-A FSS-C6 CC-I Commented [A100]: NRC-2022-0073-DRAFT-0007-1 FLESQ-D1 CC-II FES-B3 Yes FPRM-A1 Yes FSS-C7 Yes NRC-2022-0073-DRAFT-0007-2 FPRM-A2Yes NRC-2022-0073-DRAFT-0007-3 HLR-FLESQ-E HLR-FES-C FPRM-A3 YesHLR-FSS-D FLESQ-E1 Yes FES-C1 CC-I FSS-D1 Yes FLESQ-E2 CC-II FES-C2 Yes HLR-FPRM-B FSS-D2 CC-I FES-C3 Yes FPRM-B1 Yes FSS-D3 Yes HLR-FLESQ-F FPRM-B2 Yes FSS-D4 Yes FLESQ-F1 CC-II HLR-FES-D FPRM-B4 Yes FSS-D5 Yes FLESQ-F2 Yes FES-D1 Yes FPRM-B5 CC-II FSS-D6 Yes FLESQ-F3 Yes FES-D2 Yes FPRM-B6 CC-II FSS-D7 Yes FLESQ-F4 Yes FES-D3 Yes FPRM-B7 CC-II FSS-D8 Yes FLESQ-F5 Yes FPRM-B8 CC-II FSS-D9 Yes HLR-FCS-A FPRM-B9 CC-I FSS-D11 Yes Commented [A101]: NRC-2022-0073-DRAFT-0007-1 C.1.3.9 (F) FCS-A1 CC-I FPRM-B10 CC-I NRC-2022-0073-DRAFT-0007-2 HLR-FPP-A FCS-A2 Yes FPRM-B11 Yes HLR-FSS-E NRC-2022-0073-DRAFT-0007-3 FPP-A1 Yes FCS-A3 CC-I FPRM-B12 Yes FSS-E1 CC-I Commented [A102]: NRC-2022-0073-DRAFT-0007-1 FCS-A4 Yes FPRM-B13 CC-I FSS-E2 Yes NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 HLR-FPP-B F P R M - B 1 4 Ye s F S S - E 3 Ye s Commented [A103]: NRC-2022-0073-DRAFT-0007-1 FPP-B1 Yes HLR-FCS-B FPRM-B15 Ye s FSS-E4 CC-II NRC-2022-0073-DRAFT-0007-2 FPP-B2 Yes FCS-B1 Yes FPRM-B16 Yes FSS-E5 Yes NRC-2022-0073-DRAFT-0007-3 FPP-B3 Yes FCS-B2 Yes FPRM-B17 Yes Commented [A104]: NRC-2022-0073-DRAFT-0007-1 FPP-B4 Yes FCS-B3 Yes HLR-FSS-F NRC-2022-0073-DRAFT-0007-2 FPP-B6 Yes HLR-FPRM-C FSS-F1 Yes NRC-2022-0073-DRAFT-0007-3 FPP-B7 Yes HLR-FCS-C FPRM-C1 Yes FSS-F2 CC-I Commented [A105]: NRC-2022-0073-DRAFT-0007-1 FPP-B8 Ye s F C S - C 1 Ye s F P R M - C 2 Ye s NRC-2022-0073-DRAFT-0007-2 FCS-C2 Yes FPRM-C3 Yes HLR-FSS-G NRC-2022-0073-DRAFT-0007-3 HLR-FPP-C FCS-C3 Yes FPRM-C4 Yes FSS-G1 Yes Commented [A106]: NRC-2022-0073-DRAFT-0007-1 FPP-C1 Yes FSS-G2 Yes NRC-2022-0073-DRAFT-0007-2 FPP-C2 Yes HLR-FQLS-A HLR-FSS-A FSS-G3 Yes NRC-2022-0073-DRAFT-0007-3 FPP-C3 Yes FQLS-A1 Yes FSS-A1 Yes FSS-G4 CC-II Commented [A107]: NRC-2022-0073-DRAFT-0007-1 FQLS-A2 Yes FSS-A2 Yes FSS-G5 CC-II NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 HLR-FES-A FQLS-A3 Yes FSS-A3 Yes FSS-G6 CC-II FES-A1 Yes FQLS-A4 Yes FSS-A4 Yes FSS-G7 CC-II FES-A2 Yes FQLS-A5 Yes FSS-G8 Yes FES-A3 Yes FQLS-A6 Yes HLR-FSS-B FSS-G9 Yes FES-A4 Yes FSS-B1 Yes FES-A5 CC-II HLR-FQLS-B FSS-B2 CC-I HLR-FSS-H Commented [A108]: NRC-2022-0073-DRAFT-0007-1 FES-A6 CC-II FQLS-B1 Yes FSS-H1 Yes NRC-2022-0073-DRAFT-0007-2 FES-A7 Yes FQLS-B2 Yes HLR-FSS-C FSS-H2 Yes NRC-2022-0073-DRAFT-0007-3 FQLS-B3 Yes FSS-C1 CC-I FSS-H3 Yes Commented [A109]: NRC-2022-0073-DRAFT-0007-1 HLR-FES-B FSS-C2 CC-I FSS-H4 Yes NRC-2022-0073-DRAFT-0007-2 FES-B1 Yes FSS-C3 Yes NRC-2022-0073-DRAFT-0007-3 FSS-C4 CC-I Commented [A110]: NRC-2022-0073-DRAFT-0007-1 NRC-2022-0073-DRAFT-0007-2 FSS-C5 Yes NRC-2022-0073-DRAFT-0007-3

RG 1.253, App. A, Page A-16 Table A-3. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to Additional PRA Elements (3 of 6)

HLR-FIGN-A HLR-FHR-E HLR-SHA-B SHA-I2 Yes F I G N - A 1 Ye s F H R - E 1 Ye s S H A - B 1 Ye s S H A - I 3 Ye s F I G N - A 2 Ye s F H R - E 2 Ye s S H A - B 2 Ye s F I G N - A 3 Ye s F H R - E 3 Ye s S H A - B 3 Ye s H L R - S F R - A FIGN-A6 Yes SHA-B4 Yes SFR-A1 Yes FIGN-A7 CC-I HLR-FESQ-A S H A - B 5 Ye s S F R - A 2 Ye s Commented [A111]: NRC-2022-0073-DRAFT-0007-1 FIGN-A8 Yes FESQ-A1 Yes NRC-2022-0073-DRAFT-0007-2 FIGN-A9 Yes FESQ-A2 Yes HLR-SHA-C HLR-SFR-B NRC-2022-0073-DRAFT-0007-3 FIGN-A10 CC-II FESQ-A3 Yes SHA-C1 Yes SFR-B1 CC-I FIGN-A11 Yes FESQ-A4 Yes SHA-C2 Yes SFR-B2 Yes FIGN-A12 Yes FESQ-A5 CC-I S H A - C 3 Ye s S F R - B 3 Ye s Commented [A112]: NRC-2022-0073-DRAFT-0007-1 SHA-C4 Yes SFR-B4 CC-II NRC-2022-0073-DRAFT-0007-2 HLR-FIGN-B HLR-FESQ-B SHA-C5 Yes SFR-B5 CC-II NRC-2022-0073-DRAFT-0007-3 FIGN-B1 Yes FESQ-B1 Yes SFR-B6 Yes FIGN-B2 Yes HLR-SHA-D FIGN-B3 Yes HLR-FESQ-C SHA-D1 Yes HLR-SFR-C FESQ-C1 CC-I SHA-D2 Yes SFR-C1 Yes Commented [A113]: NRC-2022-0073-DRAFT-0007-1 HLR-FCF-A SHA-D3 Yes SFR-C2 Yes NRC-2022-0073-DRAFT-0007-2 FCF-A1 CC-I HLR-FESQ-D SHA-D4 Yes NRC-2022-0073-DRAFT-0007-3 FCF-A2 CC-II FESQ-D1 CC-II HLR-SFR-D FCF-A3 Yes FESQ-D2 Yes HLR-SHA-E SFR-D1 Yes FCF-A4 Yes FESQ-D3 Yes SHA-E1 Yes SFR-D2 Yes SHA-E3 Yes SFR-D4 CC-I HLR-FCF-B HLR-FESQ-E SHA-E5 Yes SFR-D5 Yes FCF-B1 Yes FESQ-E1 Yes SFR-D6 Yes FCF-B2 Yes FESQ-E2 CC-II HLR-SHA-F SFR-D7 Yes FCF-B3 Yes SHA-F1 Yes SFR-D8 Yes HLR-FESQ-F SHA-F2 Yes HLR-FHR-A FESQ-F1 Yes SHA-F3 Yes HLR-SFR-E FHR-A1 CC-I FESQ-F2 Yes SHA-F4 Yes SFR-E1 CC-I Commented [A114]: NRC-2022-0073-DRAFT-0007-1 FHR-A3 CC-II FESQ-F3 Yes SFR-E2 CC-I NRC-2022-0073-DRAFT-0007-2 FESQ-F4 Yes HLR-SHA-G SFR-E3 CC-I NRC-2022-0073-DRAFT-0007-3 HLR-FHR-B SHA-G1 Yes SFR-E4 CC-I FHR-B1 CC-II C.1.3.10 (S) SHA-G2 Yes SFR-E5 CC-I FHR-B2 CC-II HLR-SHA-A SFR-E6 Yes SHA-A1 Yes HLR-SHA-H SFR-E7 Yes HLR-FHR-C SHA-A2 Yes SHA-H1 Yes FHR-C1 CC-I SHA-A3 Yes SHA-H2 Yes HLR-SFR-F Commented [A115]: NRC-2022-0073-DRAFT-0007-1 SHA-A4 Yes SHA-H3 Yes SFR-F1 Yes NRC-2022-0073-DRAFT-0007-2 HLR-FHR-D SHA-A5 Yes SHA-H4 Yes SFR-F2 Yes NRC-2022-0073-DRAFT-0007-3 FHR-D1 Yes SHA-A6 Yes SFR-F3 Yes FHR-D2 Yes SHA-A7 Yes HLR-SHA-I FHR-D3 Yes SHA-I1 Yes HLR-SPR-A SPR-A1 Yes SPR-A2 Yes

RG 1.253, App. A, Page A-17 Table A-3. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to Additional PRA Elements (4 of 6)

SPR-A3 Ye s S P R - F 2 Ye s W H A - F 2 Ye s W F R - E 3 Ye s SPR-A4 Yes SPR-F3 Yes WHA-F3 CC-II WFR-E4 Yes SPR-F4 Yes WHA-F4 CC-II WFR-E5 Yes HLR-SPR-B SPR-F5 Yes WFR-E6 Yes SPR-B1 Yes HLR-WHA-G WFR-E7 Yes SPR-B2 Yes C.1.3.12 (W) WHA-G1 Yes WFR-E8 CC-II SPR-B3 Yes HLR-WHA-A WHA-G2 Yes WFR-E9 CC-II SPR-B4 Yes WHA-A1 Yes WHA-G3 Yes WFR-E10 Yes SPR-B5 Yes WHA-A2 Yes WFR-E11 Yes SPR-B6 Yes WHA-A3 Yes HLR-WFR-A WFR-E12 Yes SPR-B7 CC-II WHA-A4 Yes WFR-A1 Yes SPR-B8 CC-I WHA-A5 Yes WFR-A2 Yes HLR-WFR-F Commented [A116]: NRC-2022-0073-DRAFT-0007-1 SPR-B9 CC-I WHA-A6 Ye s W F R - A 3 Ye s W F R - F 1 Ye s NRC-2022-0073-DRAFT-0007-2 SPR-B10 CC-I WHA-A7 Yes WFR-A4 Yes WFR-F2 Yes NRC-2022-0073-DRAFT-0007-3 SPR-B11 CC-I WHA-A8 Yes WFR-A5 Yes Commented [A117]: NRC-2022-0073-DRAFT-0007-1 SPR-B12 CC-I WFR-A6 Yes HLR-WFR-G NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 SPR-B13 Yes HLR-WHA-B WFR-A7 Yes WFR-G1 Yes Commented [A118]: NRC-2022-0073-DRAFT-0007-1 WHA-B1 Yes WFR-A8 Yes WFR-G2 Yes NRC-2022-0073-DRAFT-0007-2 HLR-SPR-C WHA-B2 Yes WFR-A9 Yes NRC-2022-0073-DRAFT-0007-3 SPR-C1 Yes WHA-B3 Yes HLR-WFR-H Commented [A119]: NRC-2022-0073-DRAFT-0007-1 SPR-C2 Yes WHA-B4 Yes HLR-WFR-B WFR-H1 CC-I NRC-2022-0073-DRAFT-0007-2 SPR-C3 Yes WHA-B5 Yes WFR-B1 Yes WFR-H2 CC-II NRC-2022-0073-DRAFT-0007-3 SPR-C4 Yes WHA-B6 Yes WFR-B2 Yes WFR-H3 Yes Commented [A120]: NRC-2022-0073-DRAFT-0007-1 SPR-C5 Yes WFR-B3 Yes WFR-H4 Yes NRC-2022-0073-DRAFT-0007-2 SPR-C6 Yes HLR-WHA-C WFR-B4 Yes NRC-2022-0073-DRAFT-0007-3 WHA-C1 Yes WFR-B6 Yes HLR-WFR-I HLR-SPR-D WHA-C2 Yes WFR-B7 Yes WFR-I1 Yes SPR-D1 Yes WHA-C3 CC-II WFR-I2 Yes SPR-D2 CC-I WHA-C4 Yes HLR-WFR-C WFR-I3 Yes Commented [A121]: NRC-2022-0073-DRAFT-0007-1 SPR-D3 CC-II WHA-C5 Yes WFR-C1 Yes NRC-2022-0073-DRAFT-0007-2 SPR-D4 Ye s WHA-C6 Yes WFR-C2 Yes HLR-WPR-A NRC-2022-0073-DRAFT-0007-3 SPR-D5 CC-II WFR-C3 Yes WPR-A1 Yes Commented [A122]: NRC-2022-0073-DRAFT-0007-1 HLR-WHA-D WFR-C4 Yes WPR-A2 Yes NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 HLR-SPR-E WHA-D1 Yes WPR-A3 Yes SPR-E1 Yes WHA-D2 Yes HLR-WFR-D WPR-A4 Yes SPR-E2 Yes WFR-D1 Yes SPR-E3 Yes HLR-WHA-E WFR-D2 Yes HLR-WPR-B SPR-E4 Yes WHA-E1 Yes WFR-D3 Yes WPR-B1 Yes SPR-E5 CC-II WHA-E2 Yes WFR-D4 Yes WPR-B2 Yes SPR-E6 Yes WHA-E3 Yes WFR-D5 Yes WPR-B3 Yes SPR-E7 Yes WHA-E4 Yes WFR-D6 Yes WPR-B4 Yes S P R - E 8 Ye s W H A - E 5 Ye s W P R - B 5 Ye s HLR-WFR-E WPR-B6 CC-II HLR-SPR-F HLR-WHA-F WFR-E1 Yes WPR-B7 CC-I SPR-F1 Yes WHA-F1 Yes WFR-E2 Yes

RG 1.253, App. A, Page A-18 Table A-3. Non-LWR CP Applicatio ns Based on the LMP Methodology:

Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to Additional PRA Elements (5 of 6)

WPR-B8 Ye s X F H A - A 8 Yes HLR-XFFR-A HLR-XFPR-C WPR-B9 Yes XFHA-A9 Yes XFFR-A1 Yes XFPR-C1 Yes XFFR-A2 Yes XFPR-C2 Yes HLR-WPR-C HLR-XFHA-B XFFR-A3 Yes XFPR-C3 Yes WPR-C1 Yes XFHA-B1 Yes XFFR-A4 Yes XFPR-C4 Yes WPR-C2 Yes XFHA-B2 Yes XFFR-A5 Yes XFPR-C5 Yes WPR-C3 Yes XFHA-B3 Yes XFPR-C6 CC-I Commented [A123]: NRC-2022-0073-DRAFT-0007-1 WPR-C4 Yes XFHA-B4 Yes HLR-XFFR-B XFPR-C7 Yes NRC-2022-0073-DRAFT-0007-2 WPR-C5 Yes XFFR-B1 Yes XFPR-C8 Yes NRC-2022-0073-DRAFT-0007-3 HLR-XFHA-C XFFR-B3 Yes XFPR-C9 Yes HLR-WPR-D XFHA-C1 Yes XFFR-B4 Yes XFPR-C10 Yes WPR-D1 Yes XFHA-C2 CC-II XFFR-B5 Yes XFPR-C11 CC-II Commented [A124]: NRC-2022-0073-DRAFT-0007-1 WPR-D2 Yes XFHA-C3 CC-II XFPR-C12 Yes NRC-2022-0073-DRAFT-0007-2 WPR-D3 CC-I XFHA-C4 Yes HLR-XFFR-C NRC-2022-0073-DRAFT-0007-3 WPR-D4 CC-II XFHA-C5 Yes XFFR-C1 CC-I HLR-XFPR-D Commented [A125]: NRC-2022-0073-DRAFT-0007-1 WPR-D6 Yes XFHA-C6 Yes XFFR-C2 Yes XFPR-D1 Yes NRC-2022-0073-DRAFT-0007-2 WPR-D7 Ye s XFHA-C7 Yes XFPR-D2 Yes NRC-2022-0073-DRAFT-0007-3 WPR-D8 Yes XFHA-C8 Yes HLR-XFFR-D XFPR-D3 Yes Commented [A126]: NRC-2022-0073-DRAFT-0007-1 NRC-2022-0073-DRAFT-0007-2 WPR-D9 Yes XFHA-C9 Yes XFFR-D1 CC-I XFPR-D4 Yes NRC-2022-0073-DRAFT-0007-3 WPR-D10 Yes XFHA-C10 Yes XFFR-D2 CC-II XFPR-D5 Yes Commented [A127]: NRC-2022-0073-DRAFT-0007-1 WPR-D11 CC-I XFHA-C11 Yes XFFR-D3 Yes NRC-2022-0073-DRAFT-0007-2 XFFR-D4 Yes HLR-XFPR-E NRC-2022-0073-DRAFT-0007-3 HLR-WPR-E HLR-XFHA-D XFPR-E1 Yes WPR-E1 Yes XFHA-D1 CC-II HLR-XFFR-E XFPR-E2 CC-I Commented [A128]: NRC-2022-0073-DRAFT-0007-1 WPR-E2 Yes XFHA-D2 Yes XFFR-E1 Yes XFPR-E3 CC-II NRC-2022-0073-DRAFT-0007-2 WPR-E3 Yes XFHA-D3 CC-I XFFR-E2 Yes XFPR-E4 Ye s NRC-2022-0073-DRAFT-0007-3 W P R - E 4 Ye s X F H A - D 4 Ye s X F P R - E 5 Ye s Commented [A129]: NRC-2022-0073-DRAFT-0007-1 WPR-E5 CC-II HLR-XFFR-F XFPR-E6 CC-I NRC-2022-0073-DRAFT-0007-2 WPR-E6 Yes HLR-XFHA-E XFFR-F1 Yes XFPR-E7 Yes NRC-2022-0073-DRAFT-0007-3 WPR-E7 Yes XFHA-E1 Yes XFFR-F2 Yes XFPR-E8 Yes Commented [A130]: NRC-2022-0073-DRAFT-0007-1 XFHA-E2 Yes XFFR-F3 Yes NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 HLR-WPR-F XFHA-E3 CC-II HLR-XFPR-F WPR-F1 Yes XFHA-E4 CC-II HLR-XFPR-A XFPR-F1 Yes WPR-F2 Yes XFPR-A1 Yes XFPR-F2 Yes WPR-F3 Yes HLR-XFHA-F XFPR-A2 Yes XFPR-F3 Yes XFHA-F1 Yes XFPR-A3 Yes XFPR-F4 Ye s Commented [A131]: NRC-2022-0073-DRAFT-0007-1 C.1.3.13 (XF) XFHA-F2 Yes XFPR-A4 Yes XFPR-F5 CC-II NRC-2022-0073-DRAFT-0007-2 HLR-XFHA-A XFHA-F3 Yes XFPR-A5 Yes XFPR-F6 Yes NRC-2022-0073-DRAFT-0007-3 XFHA-A1 Yes XFHA-F4 Yes XFPR-A6 Yes XFPR-F7 Yes XFHA-A2 Yes XFPR-A7 CC-II XFHA-A3 Yes HLR-XFHA-G HLR-XFPR-G XFHA-A4 Yes XFHA-G1 Yes HLR-XFPR-B XFPR-G1 Yes X F H A - A 5 Ye s X F H A - G 2 Ye s X F P R - B 1 Ye s X F P R - G 2 Ye s X F H A - A 6 Ye s X F H A - G 3 Ye s X F P R - B 2 Ye s X F P R - G 3 Ye s XFHA-A7 Yes XFPR-B3 Yes

Table A-3. Non-LWR CP Applicatio ns Based on the LMP Methodology:

RG 1.253, App. A, Page A-19 Applicability of ASME/ANS Non-LWR PRA Standard High-Level Requi rements and Supporting Requirements to Additional PRA Elements (6 of 6)

HLR-XFPR-H HLR-OPR-B XFPR-H1 Yes OPR-B1 Yes XFPR-H2 Yes OPR-B2 Yes XFPR-H3 Yes OPR-B3 Yes OPR-B4 Yes C.1.3.14 (O) OPR-B5 Yes HLR-OHA-A OPR-B6 Yes OHA-A1 Yes OPR-B7 Yes OHA-A2 Yes OPR-B8 CC-II OHA-A3 CC-II OPR-B9 CC-I Commented [A132]: NRC-2022-0073-DRAFT-0007-1 OHA-A4 CC-I OPR-B10 CC-I NRC-2022-0073-DRAFT-0007-2 OHA-A5 Yes OPR-B11 CC-I NRC-2022-0073-DRAFT-0007-3 OHA-A6 Yes OPR-B12 CC-I Commented [A133]: NRC-2022-0073-DRAFT-0007-1 OHA-A7 Yes NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 OHA-A8 CC-I HLR-OPR-C Commented [A134]: NRC-2022-0073-DRAFT-0007-1 OHA-A9 Yes OPR-C1 Yes NRC-2022-0073-DRAFT-0007-2 OHA-A10 Yes OPR-C2 CC-I NRC-2022-0073-DRAFT-0007-3 OPR-C4 Yes Commented [A135]: NRC-2022-0073-DRAFT-0007-1 HLR-OHA-B OPR-C5 Yes NRC-2022-0073-DRAFT-0007-2 OHA-B1 Yes OPR-C6 CC-I NRC-2022-0073-DRAFT-0007-3 OHA-B2 Yes OPR-C7 Yes Commented [A136]: NRC-2022-0073-DRAFT-0007-1 OHA-B3 Yes OPR-C8 Yes NRC-2022-0073-DRAFT-0007-2 NRC-2022-0073-DRAFT-0007-3 HLR-OFR-A HLR-OPR-D OFR-A1 CC-I OPR-D1 Yes OFR-A2 Ye s O P R - D 2 Ye s OFR-A3 Yes OPR-D3 Yes OFR-A4 CC-I OPR-D4 Yes OFR-A5 Yes OPR-D5 Yes OFR-A6 Yes OPR-D6 CC-II OFR-A7 Yes OPR-D7 Yes OPR-D8 Yes HLR-OFR-B OPR-D9 Yes OFR-B1 Yes OFR-B2 Yes HLR-OPR-E OFR-B3 Yes OPR-E1 Yes OPR-E2 Yes HLR-OPR-A OPR-E3 Yes OPR-A1 Yes OPR-E4 Yes OPR-A2 Yes OPR-E5 Yes OPR-A3 Yes OPR-A4 Yes

RG 1.253, App. A, Page A-20 Acronyms/Abbreviations

ADAMS Agencywide Documents Access and Management System ANS American Nuclear Society ASME American Society of Mechanical Engineers CC Capability Category CFR Code of Federal Regulations CP construction permit DBHL design-basis hazard level DBEHL design-basis external hazard level EAB exclusion area boundary ER environmental report EPZ emergency planning zone FR Federal Register LBE licensing basis event LMP Licensing Modernization Project NEI Nuclear Energy Institute non-LWR non-light-water reactor NRC Nuclear Regulatory Commission OL operating license POS plant operating state PRA probabilistic risk assessment PSAR preliminary safety analysis report QHO quantitative health objective RG regulatory guide RIM reliability and integrity management RSF required safety function SAMA severe accident mitigation alternative SAMDA severe accident mitigation design alternative SMR small modular reactor SR SSCs safety-related systems, structures, and components

RG 1.253, App. A, Page A-21 References6

A-1 Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Domestic Licensing of Production and Utilization Facilities.

A-2 Nuclear Energy Institute (NEI), NEI 1804, Revision 1, Ris k-Informed Performance-Based Technology-Inclusive Guidance for Non-Light-Water Reactor Licen sing Basis Development, Washington, DC, August 2019. (Agencywide Documents Access and M anagement System (ADAMS) Accession No. ML19241A472)

A-3 Nuclear Regulatory Commission (NRC), RG 1.247 for trial use, Acceptability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor Risk-Informed Activities, Washington, DC.

A-4 NEI, NEI 21-07, Revision 1, Technology Inclusive Guidance for Non-Light Water Reactors, Safety Analysis Report Content: For Applicants Using the NEI 18 -04 Methodology, Washington, DC, February 2022. (ML22060A190)

A-5 NRC, RG 1.233, Guidance for a Technology-Inclusive, Risk-I nformed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applic ations for Licenses, Certifications, and Approvals for Non-Light-Water Reactors, Wa shington, DC.

A-6 American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS)

RA-S-1.4-2021, Probabilistic Risk Assessment Standard for Adva nced Non-Light Water Reactor Nuclear Power Plants, New York, NY, 2021.

A-7 NRC, DANU-ISG-2022-01, Review of Risk-Informed, Technology -Inclusive Advanced Reactor Applications-Roadmap, Washington, DC (ML23277A139)

A-8 NRC, NUREG-2122, Glossary of Risk-Related Terms in Support of Risk-Informed Decisionmaking, Washington, DC, November 2013.

A-9 NRC, RG 4.2, Preparation of Environmental Reports for Nucl ear Power Stations, Washington, DC.

6 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 (PDR) staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC 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.

Publications from the Nuclear Energy Institute (NEI) are availa ble at their Web site: http://www.nei.org/ or by contacting the headquarters at Nuclear Ene rgy Institute, 1776 I Street NW, Washington DC 20006-3708; telephone: 202-739-800; fax 202-785-4019.

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

RG 1.253, App. A, Page A-22 A-10 NRC, Policy Statement on the Regulation of Advanced React ors, Federal Register, Vol. 73, No.

199, October 14, 2008, pp. 60612-60616 (73 FR 60612).

A-11 NRC, Safety Goals for Operations of Nuclear Power Plants; Policy Statement; Republication, Federal Register, Vol. 51, No. 162, August 21, 1986, pp. 30028-30033 (51 FR 300 28).

A-12 NRC, Severe Reactor Accidents Regarding Future Designs an d Existing Plants, Federal Register, Vol. 50, No. 153, August 8, 1985, pp. 32138-32150 (50 FR 3213 8).

A-13 NRC, Use of Probabilistic Risk Assessment Methods in Nucl ear Regulatory Activities, Federal Register, Vol. 60, No. 158, August 16, 1995, pp. 42622-42629 (60 FR 426 22).

A-14 NRC, DANU-ISG-2022-07, Risk-Informed Inservice Inspection /Inservice Testing Programs for Non-LWRs, Washington, DC (ML23277A145)

A-15 ASME, Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 2, Requirements for Reliabil ity and Integrity Management (RIM) Programs for Nuclear Power Plants, 2019 Edition, New York, NY, July 1, 2019.

A-16 NRC, RG 1.246, Acceptability of ASME Code,Section XI, Di vision 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Powers Plants, for Non-Light-Water Reactors, Washington, DC.

A-17 NRC, DANU-ISG-2022-08, Risk-Informed Technical Specificat ions, Washington, DC.

(ML23277A146)

A-18 NRC, DANU-ISG-2022-09, Risk-Informed, Performance-Based F ire Protection Program (for Operations), Washington, DC. (ML23277A147)

A-19 NRC, NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Washington, DC, March 20 17.

A-20 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Washington, DC.

A-21 NRC, DANU-ISG-2022-05, Organization and Human-System Cons iderations, Washington, DC. (ML23277A143)

A-22 NEI, NEI 2009, Revision 1, Performance of PRA Peer Reviews Using the ASME/ANS Advanced Non-LWR PRA Standard, Washington, DC, May 2021. (ML21125A284)

RG 1.253, App. A, Page A-23