ML22143A805

NRR E-mail Capture - NRC Initial Comments on Technology Inclusive Risk-Informed Change Evaluation (Tirice) White Paper

ML22143A805
Person / Time
Issue date:	05/20/2022
From:	Michael Orenak NRC/NRR/DANU
To:	Afzali A, Tschiltz M Southern Company Services
References
N-2022-ADV-0004
Download: ML22143A805 (41)

Text

From: Orenak, Michael Sent: Friday, May 20, 2022 3:32 PM To: michael.tschiltz@gmail.com; AAFZALI@southernco.com

Subject:

NRC initial comments on TIRICE white paper Attachments: NRC initial comments for TIRICE Project White Paper Rev A 5-20-22.docx

Amir Afzali Southern Company Services Licensing and Policy Director - Next Generation Reactors

Michael Tschiltz Southern Company Services

Mr. Afzali and Mr. Tschiltz,

The purpose of this email is to provide you with the NRCs initial comments on the attached document titled, Technology Inclusive Risk Informed Change Evaluation (TIRICE) For Non-Light Water Reactors; Change Control Scope and Process For a Reactor Licensed in Accordance with the NEI 18-04 Guidance. The attached document with NRC comments will be discussed during the planned June 9, 2022, public meeting on this document. The meeting notice and agenda for this meeting will be issued next week.

Please note that these comments are preliminary given that the requested fast turnaround and an understanding that this document will be undergoing further revision in the near term.

This email will be captured in ADAMS. Both the email and document with comments will be made publicly available so that interested stakeholders will have access to the information for the June 9th public meeting.

If you have any immediate questions regarding the attached document that wont be discussed in the June 9th public meeting, please contact me.

Sincerely,

Mike Orenak

Michael D. Orenak, Project Manager Advanced Reactor Licensing Branch 1 (UAL1)

NRR - Division of Advanced Reactors and Non-Power Production and Utilization Facilities (DANU) 301-415-3229 Michael.Orenak@nrc.gov

Hearing Identifier: NRR_DRMA Email Number: 1645

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NRC initial comments on TIRICE white paper Sent Date: 5/20/2022 3:31:49 PM Received Date: 5/20/2022 3:31:00 PM From: Orenak, Michael

Created By: Michael.Orenak@nrc.gov

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"michael.tschiltz@gmail.com" <michael.tschiltz@gmail.com>

Tracking Status: None "AAFZALI@southernco.com" <AAFZALI@southernco.com>

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Technology Inclusive Risk Informed Change Evaluation (TIRICE)

For Non-Light Water Reactors

Change Control Scope and Process For a Reactor Licensed in Accordance with the NEI 18-04 Guidance

Document Number SC-16166-107 Revision A

Battelle Energy Alliance, LLC Contract No. 221666 SOW-16166

May 2022

Prepared for:

U.S. Department of Energy (DOE)

Office of Nuclear Energy Under DOE Idaho Operations Office Contract DE-AC07-05ID14517

Technology Inclusive Risk Informed Change Evaluation (TIRICE)

For Non-Light Water Reactors

Change Control Scope and Process For a Reactor Licensed in Accordance with the NEI 18-04 Guidance

Document Number SC-16166-107 Revision A

Battelle Energy Alliance, LLC Contract No. 221666 SOW-16166

Issued for Collaborative Review by:

Amir Afzali, Next Generation Licensing and Policy Director Date Southern Company Services

TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

Disclaimer

This report was prepared as an account of work sponsored by an agency of the United States (U.S.) Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, nor Southern Company, Inc., nor any of its employees, nor any of its subcontractors, nor any of its sponsors or co-funders, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

Abstract

Nuclear Regulatory Commission (NRC) regulation 10 CFR 50.59 establishes criteria for determining if prior NRC approval is required before implementing changes to a reactor licensed under 10 CFR Part 50 or 10 CFR Part 52. Nuclear Energy Institute document NEI 96-07 Guidelines for 10 CFR 50.59 Implementation provides guidance for applying the 10 CFR 50.59 criteria to currently-operating light water reactors (LWRs). This paper provides supplemental 10 CFR 50.59 guidance for advanced non-LWRs that were licensed using the Commented [A1]: Can SMRs or large LWRs that use LMP use methodologies in NEI 18-04 Risk-Informed Performance-Based Technology Inclusive TIRICE?

Guidance for Non-Light Water Reactor Licensing Basis Development and NEI 21-07 Technology Inclusive Guidance for Non-Light Water Reactors - Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology. Used in conjunction with NEI 96-07, this guidance should allow non-LWR licensees to implement appropriate change control programs for the operation of their reactors.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

Table of Contents

1.0 Introduction..................................................................................................................................... 1

1.1 Purpose and Scope

................................................................................................................. 1 1.2 Regulatory Approach.............................................................................................................. 2 1.3 Background............................................................................................................................ 2 NEI 96-07 Guidelines for 10 CFR 50.59 Implementation.......................................... 2 NEI 18-04 Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development............................................ 4 NEI 21-07 Technology Inclusive Guidance for Non-Light Water Reactors - Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology............... 5 1.4 Application of this Guidance................................................................................................... 6 2.0 NEI 96-07 Introductory Material...................................................................................................... 7 2.1 Introduction (NEI 96-07 Section 1.0)....................................................................................... 7 2.2 Defense-in-Depth Design Philosophy and 10 CFR 50.59 (NEI 96-07 Section 2.0)...................... 7 2.3 Definitions and Applicability of Terms (NEI 96-07 Section 3.0)................................................ 8 Accident Previously Evaluated in the FSAR (as Updated) (NEI 96-07 Section 3.2)........ 8 Change (NEI 96-07 Section 3.3).................................................................................. 8 Malfunction of an SSC Important to Safety (NEI 96-07 Section 3.9)............................ 8 Safety Analyses (NEI 96-07 Section 3.12).................................................................... 9 3.0 Implementation Guidance............................................................................................................. 10 3.1 Applicability......................................................................................................................... 11 Applicability to Licensee Activities (NEI 96-07 Section 4.1.1).................................... 12 Maintenance Activities (NEI 96-07 Section 4.1.2)..................................................... 12 UFSAR Modifications (NEI 96-07 Section 4.1.3)........................................................ 12 Changes to Procedures Governing the Conduc t of Operations (NEI 96-07 Section 4.1.4)........................................................................................................... 12 Changes to Approved Fire Protection Programs (NEI 96 -07 Section 4.1.5)................ 12 Changes to the Probabilistic Risk Assessment (PRA)................................................. 12 Changes to the State of Knowledge.......................................................................... 13 3.2 Screening............................................................................................................................. 13 Change to the Facility or Procedures (NEI 96-07 Section 4.2.1)................................. 14 Changes to the Facility as Described in the UFSAR (NEI 96 -07 Section 4.2.1.1)......... 14 Changes to Procedures as Described in the UFSAR (NEI 96 -07 Section 4.2.1.2)......... 14 Changes to UFSAR Methods of Evaluation ( NEI 96-07 Section 4.2.1.3)..................... 14 Test or Experiment Not Described in the UFSAR (NEI 96 -07 Section 4.2.2)............... 15 3.3 Evaluation............................................................................................................................ 15

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

Evaluation Criteria................................................................................................... 15 Evaluation Process................................................................................................... 20 4.0 Documentation and Reporting (NEI 96-07 Section 5.0)................................................................... 24 5.0 Summary....................................................................................................................................... 25

Appendix A Probabilistic Risk Assessment............................................................................................ A-1

Appendix B Terminology and Definitions............................................................................................. B-1

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

List of Abbreviations

AOO Anticipated Operational Occurrence ARCAP Advanced Reactor Content of Application Project CFR Code of Federal Regulations DID Defense-in-Depth FSAR Final Safety Analysis Report LBE Licensing Basis Event LMP Licensing Modernization Project LWR Light water reactor NEI Nuclear Energy Institute non-LWR Non-light water reactor NRC Nuclear Regulatory Commission NSRST Non-Safety-Related with Special Treatment NUREG Nuclear Regulatory Commission technical report designation PDC Principal Design Criteria PRA Probabilistic Risk Assessment SAR Safety Analysis Report SR Safety-Related SSCs Structures, Systems, and Components UFSAR Updated Final Safety Analysis Report

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

1.1 Purpose and Scope

The purpose of this paper is to describe a proposed process for determining if prior regulatory approval is necessary for changes to certain advanced non-LWRs licensed for power production and/or other uses under 10 CFR Part 50 or 10 CFR Part 52. The process is applicable only to Commented [A2]: Certified designs have their own 50.59-like non-light water reactor licensees that implemented NEI 18-04, Risk-Informed Performance-change process. Conceivably, a COL referencing a DC may need to use both the 50.59 and 50.59-like change process depending on the Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis source of the information in its SAR.

Development, consistent with Regulatory Guide 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content Does this paper include a new change process for DCs using the LMP process?

of Applications for Licenses, Certifications, and Approvals for Non-Light Water Reactors. The Commented [A3]: Is this process forward looking such that it NEI 18-04 methodology is also referred to as the Licensing Modernization Project (LMP) could be the same change process for OLs and COLs issues under methodology. Part 53? It will be confusing to have guidance for change process for OLs and Custom COLs, vs DCs, VS licenses under 53. Could one process be developed that works and can be implemented on (1) 10 CFR 50.59 (also applicable to 10 CFR Part 52) permits licensees to make changes to the DCs via DC rulemakings, (2) COLs and OLs via exemption to 50.59 facility without prior Nuclear Regulatory Commission (NRC) approval, provided the and (3) for licenses under 53 with Part 53 rule, with one consistent process for implementation?

requirements in the regulation are met. Change control guidance is mature and in place for Commented [A4]: There is limited applicability based on currently-operating LWRs. However, the existing change control guidance is tailored for the whether a COL references a DC as DCs have their own change physical characteristics of LWRs and the terminology and approach of a traditiona l, process. Either the scope of this white paper should be limited to deterministically-derived safety case. Advanced non-LWRs may elect to follow NEI 18 -04 for just OLs under Part 50 (or custom COLs under Part 52 that do not reference a certified design) or it should include much more selection of licensing basis events (LBEs); safety classification of structures, systems, and discussion on the 50.59-like change process included in the DC rules components (SSCs) and associated special treatments; and determinat ion of Defense-in-Depth or what a 50.59-like change process would look like for an LMP-(DID) adequacy. The resulting LMP-based affirmative safety case is substantially different from based DC.

the traditional deterministic, compliance-based safety cases in place for LWRs licensed by NRC. Commented [A5]: 1)Change control for COLs referencing certified design information has had mixed results. Suggest The attributes of the LMP-based affirmative safety case require additional guidance for efficient obtaining specific feedback from Southern Company Vogtle 3&4 application of 10 CFR 50.59. staff on how this worked and any lessons learned.

2) exploring any lessons from any NRC inspections of the 50.59 The objectives of this guidance include: and 50.59-like processes for Vogtle 3 and 4 may be valuable (if they exist).

• Provide regulatory confidence that the threshold for regulatory review of changes to the

3) Look at the Part 50/52 alignment rule because there possibly facility as described in the licensing basis will be effectively established and efficiently are proposed changes to the change process contained within managed that may need to be accounted for.

Commented [A6]: Note: This terminology is not being used in

• Minimize the unnecessary burden to the regulator and operators for determining if the TICAP Draft Guide (DG-1404).

changes require a license amendment

• Establish a clear understanding and process for how the criteria for making changes to the facility as described in the licensing basis without prior NRC approval may be met

Reactors Licensees that follow NEI 18-04 are also expected to conform to NEI 21-07, Technology Inclusive Guidance for Non-Light Water Reactors - Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology. The NEI 18-04 methodology relies on information from a comprehensive Probabilistic Risk Assessment (PRA), and the NEI 21-07 guidance anticipates that the PRA will conform to ANSI/ASME/ANS RA-S-1.4-2021, Probabilistic Risk Assessment Standard for Advanced non-Light Water Reactor Nuclear Power Plants (referred to herein as the Non-LWR PRA Standard). The guidance in this white paper

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

applies to licensees who follow NEI 18-04, NEI 21-07, and the Non-LWR PRA Standard.

Licensees that deviate from elements of NEI 18-04, NEI 21-07, or the Non-LWR PRA Standard must justify the application of this guidance to change contro l. The NRC is currently developing draft guidance in DG-1404 to endorse NEI 21-07, Revision 1, with additions and clarifications.

In addition, the NRC has issued for trial use RG 1.247 which endorses the use of the ANSI/ASME/ANS RA-S-1.4-201.

10 CFR 50.59 is only one of many processes that apply to nuclear power reactors. The regulation addresses the need for prior NRC approval for certain changes to a facility that is licensed under 10 CFR Part 50 or 10 CFR Part 52. Other regulatory processes address areas such as operability, reportability, corrective action, and changes in the state of knowledge.

Regulatory Approach Commented [A7]: The Part 53, Subpart I, change control process is very high level. Would it be better to work with the Part 53 words in lieu of the detail in TIRICE?

1.2

At this point, two options are being considered for the incorporation of this guidance into the Commented [A8]: These two options may or may not be regulatory framework. The first option is to utilize this guidance to interpret the application of feasible. They still have to be reviewed by NRC OGC before any NRC endorsement.

10 CFR 50.59 for advanced non-LWRs that were licensed using the methodologies in NEI 18-04. This supplemental guidance would be used in conjunction with existing guidance in NEI 96-07 to comply with the existing 10 CFR 50.59 regulation. This approach should allow Commented [A9]: Either address limitations of applicability to advanced non-LWR licensees to implement appropriate change control programs for the OLs or expand to address considerations for 50.59-like change process included in DC rules (or what such a 50.59-like change operation of their reactors, and it would require no additional enabling regulatory actions. The process would look like for an LMP-based DC).

second option is functionally equivalent to the first, i.e., to use this guidance in conjunction with Commented [A10]: Since Option 2 would likely require an the existing guidance in NEI 96-07 for implementation of change control programs. However, exemption, why wouldn't Option 1 also require an exemption?

the second option would be invoked by a condition that can be incorporated into the operating license, likely coupled with an exemption under 10 CFR 50.12 to the applicability, in whole or in part, of 10 CFR 50.59. Recommendations on the regulatory approach will be provided outside of Commented [A11]: Perhaps a clarification here would help as this paper and following discussions with NRC. to whether this statement means that a specific license condition could be included that provides a change process in a similar manner to the one included for changes to Fire Protection program 1.21.3 Background in accordance with GL 96-03 on operating plants.

NEI 96-07 Guidelines for 10 CFR 50.59 Implementation

10 CFR 50.59 is a lynchpin in the current regulatory framework supporting the operation of the nuclear power plant fleet. It determines the regulatory threshold for when NRC must review and approve a proposed change to the facility before its implementation.

Expanding upon that purpose, 10 CFR 50.59 is not a determination of safety nor of overall acceptability. It defines the boundary between those proposed activities changes to the facility that can be implemented by the licensee without prior NRC approval and those that must receive NRC review and approval before implementation. Commented [A12]: At some point, we may want to revisit how this is defined. For operating plants, the change could be made but The 10 CFR 50.59 rule was initially promulgated in 1962. However, by 1999 numerous not declared operable (i.e., implemented). This practice and philosophy was upset by the changes during construction license opportunities for improvement had been identified. As such, the regulation underwent a major condition for new reactors (see Vogtle 3 license condition 2.D.(1),

revision in 2000. The purposes of this revision were to: Changes During Construction, for an example). The staff notes that the Part 50/52 rule is considering rule language changes in this area and that TIRICE may need to eventually align with the outcome of this rule.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

• Establish clear definitions to promote a common understanding of the rules requirements

• Clarify the criteria for determining when changes, tests, and experiments require prior NRC approval

• Provide greater flexibility to licensees, primarily by allowing changes that have minimal safety impact to be made without prior NRC approval

• Clarify the threshold for screening out changes that do not require full evaluation under 10 CFR 50.59, primarily by the adoption of key definitions

Significant changes to the regulation included clarification of many fundamental concepts, insertion of the word minimal into the evaluation of impacts, and incorporation of the concept of screening into the regulation.

In order to ensure effective and consistent implementation of this expansive change, NRC Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, And Experiments, was issued in 2000 and linked to the rules implementation. In addition, NRC endorsed NEI 96-07, an industry guidance document addressing change control, and focused on the 2000 revision of 10 CFR 50.59. Commented [A13]: Note that RG 1.187 is currently as Revision 3, issued June 2021. It was revised to endorse an updated appendix NEI 96-07 provides detailed guidance for the three major sub-processes that comprise the larger providing guidance on applying 50.59 to digital I&C modifications and included clarifications to HFE screening examples.

10 CFR 50.59 process as it applies to LWRs. These sub-processes are applicability determination, screening, and evaluation.

The applicability determination sub-process addresses a provision of the 2000 revision that excludes proposed changes controlled by other, more specific regulations. This provision ensures that 10 CFR 50.59 is applied to proposed activities for which it is suited and allows the entire spectrum of regulations to more effectively control other activities. As an example, consistent with this provision, 10 CFR 50.59 would not be applied to any aspect of corrective action. Commented [A14]: Another more common example is changes to Technical Specifications. A license amendment under 50.90 is The screening sub-process provides for an upfront determination that an activity has no potential required to change TS rather than via 50.59. Another example is security changes under 50.54(p).

for requiring prior NRC review and approval. Activities that are screened out do not have to undergo the more resource-intensive evaluation process. Commented [A15]: This discussion could be expended to include some of the typical questions licensees have included in The evaluation sub-process is a more detailed review and evaluation of proposed activities that their screening processes (or refer to a particular section in RG 1.187 on the screening process).

screen in. The evaluation sub-process implements the 10 CFR 50.59(c)(2) criteria for evaluating the need for prior NRC review and approval for an activity. It involves addressing specific questions associated with the licensing basis for the facility and is structured around the licensing framework as described below.

As defined at the outset, 10 CFR 50.59 defines a regulatory threshold for obtaining prior NRC review and approval of proposed changes. As such, its structure replicates the licensing framework of the affected facilities. Specifically, this means that 10 CFR 50.59 is oriented around preserving these three licensing fundamentals:

1. The assumptions concerning the initiation, both frequency and type, of design basis events

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

2. The reliability and effectiveness of the mitigation systems Commented [A16]: This should be more than just mitigation -

also prevention of transients/accidents (for SSCs important to

3. The acceptability of consequences (dose) by limiting increases in the dose results of the safety) and fission product barrier performance (i.e.,

50.59(b)(2)(viii)).

postulated design basis events Commented [A17]: Is it clear to all stakeholders that applying the proposed TIRICE approach addresses all LBEs (including BDBEs),

NEI 96-07 also has five appendices attached to the base document. A summary of each is thereby expanding this fundamental scope?

provided below.

• Appendix AThe appendix consists of the text of 10 CFR 50.59.

• Appendix BThis appendix addresses the application of an analogous regulation for independent spent fuel storage installations (10 CFR 72.48). Appendix B has been superseded by NEI 12-04, Guidelines for 10 CFR 72.48 Implementation.

• Appendix CThis appendix provides guidance for applying 10 CFR 50.59 to facilities licensed under 10 CFR 52. While this appendix has not been formally endorsed, Commented [A18]: Not sure this is accurate. The acceptable Regulatory Guide 1.187 now states that Appendix C is acceptable for use by licensees for use by licensees in the letter back to NEI is sufficient.

during formal NRC endorsement via the NRCs regulatory guide process.

• Appendix DThis appendix provides very specific guidance for applying 10 CFR 50.59 to digital modifications. This guidance builds upon the guidance contained in NEI 96-07 and is intended to be used in conjunction with the base document. Appendix D was endorsed in Revision 2 of Regulatory Guide 1.187 in June 2020.

• Appendix EThis appendix provides user guidance for 16 specific situations that are commonly encountered. It uses existing guidance from NEI 96-07 to address these situations. The appendix has not been formally endorsed by NRC.

NEI 18-04 Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development

NEI 18-04 Revision 1 (August 2019) presents a technology-inclusive, risk-informed, and performance-based process for selection of LBEs; safety classification of SSCs and associated risk-informed special treatments; and determination of DID adequacy for non-LWRs including, but not limited to, molten salt reactors, high-temperature gas cooled reactors, and a variety of fast reactors at all thermal power capacities. NRC endorsed the methodology in Regulatory Guide 1.233 Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors (June 2020). Commented [A19]: ARCAP and TICAP are including LWR-SMRs since they are included in the NEIMA definition of advanced Significant attributes of the methodology that relate to the application of change control are reactors. There needs to be a reconciliation and recognition made about use of LMP for LWR-SMRs and then applicability of the LMP-summarized below. based 50.59 change process to LWR-SMRs.

• PRA plays a central role in the identification of LBEs, quantification of their frequency and consequences, and evaluation of their risk significance.

• LBEs consist of anticipated operational occurrences (AOOs), design basis events (DBEs),

beyond design basis events (BDBEs), and design basis accidents (DBAs). AOOs, DBEs, and BDBEs are composed of event sequence families identified and evaluated in the PRA.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

• DBAs are defined using a set of deterministic rules that include the identification of Required Safety Functions that are needed to keep DBEs and high consequence BDBEs inside a Frequency-Consequence (F-C) Target. DBAs are derived from DBEs but rely Commented [A20]: This text is associated with the definition of upon only Safety-Related SSCs for performance of the Required Safety Functions, and the safety-related. Including BDBEs in the bullet may create confusion regarding the DBA discussion thats the focus of the bullet.

DBAs are evaluated deterministically with consequences compared against the same dose limits applied to LWR DBAs. Suggest deleting the highlighted text.

• The remaining LBEs (AOOs, DBEs, and BDBEs) are evaluated realistically as part of the Commented [A21]: 1)It is possibly worth discussing somewhere in this paper the options that applicants have under PRA. LMP to develop source terms. Some may use the traditional approach whereas others may use the mechanistic source term

• A systematic process is used to ensure that plant capabilities and programs are sufficient to even within the context of a deterministic analysis - so, may not enable SSCs and associated human actions to perform safety-significant functions that be exactly like SRP Chapter 15 accident analysis.

2) The DBA is more deterministic/conservative but derived from provide adequate DID. the risk-informed LBEs and its a scenario specific mechanistic source term.

• Light water reactor general design criteria from 10 CFR 50 Appendix A, including the 3)They are dose criteria, not limits single failure criterion, are not imposed on the design. DID adequacy assessment along Commented [A22]: Realistically is not clear, so suggest with Reliability and capability targets are used in lieu of the single failure criterion to different wording, such as using realistic assumptions and inputs or other more clearly defined terms.

ensure that SSCs and supporting human actions provide reasonable assurance of adequate Commented [A23]: 1)This statement could be problematic if protection of public safety. Therefore, redundancy and diversity are not required (in the LWR-SMRs are included in the scope of this white paper. See traditional sense), but they may be used by the designer (on a functional basis) to meet the previous comment about inclusion of LWR-SMRs in the scope of performance targets. ARCAP and TICAP.

2) Note, if a LWR desires to use LMP, the applicant likely would need to propose PDCs (instead of the GDC) and use the NEI 18-04 addresses how to establish an LMP-based safety case for an advanced non-LWR. exemption process That safety case becomes part of the licensing basis of the reactor when NRC issues a Commented [A24]: Suggest that Southern Company clarify this 10 CFR Part 50 or Part 52 operating license for the reactor (or certifies the design under Part 52). statement to reflect that LMP considers both single and multiple failures. As currently written, and external stakeholder could come Nothing in the guidance described in this white paper affects the substance of that initial LMP-away with the impression that LMP-based designs are not single based safety case. This guidance applies only to activities that take place subsequent to initial failure tolerant.

licensing which may involve changes that impact the safety caselicensing basis. (Note that this topic is addressed and more clearly described in Section 3.3.2 on Category 1 Accidents.)

NEI 21-07 Technology Inclusive Guidance for Non-Light Water Reactors - Commented [A25]: See previous comment in Section 1.1 about Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology the use of safety case. It is not terminology that NRC uses. NRC uses safety analysis and licensing basis. This will be made more clear in DG-1404.

NEI 21-07 Revision 1 (February 2022) describes one acceptable means of developing portions of the Safety Analysis Report (SAR) content for advanced reactor applicants that utilize NEI 18-04.

The guidance describes eight chapters of a non-LWR SAR related directly to the implementation of the NEI 18-04 methodology. The chapters do not follow the standard LWR SAR outline as provided in NUREG-0800 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition. The intent of the guidance is to help ensure completeness of information submitted to NRC while avoiding unnecessary burden on the applicant and rightsizing the content of the application commensurate with the complexity of the design being reviewed.

Significant attributes of the methodology that relate to the application of change control are summarized below.

• The document describes the LMP-based affirmative safety case which is developed through the application of the NEI 18-04 methodology.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

• Applicants are expected to describe the PRA at a summary level and provide key results related to the LMP-based affirmative safety case.

• DBA analyses are documented in the SAR consistent with LWR DBAs.

• AOOs, DBEs, and BDBEs are also documented in the SAR, but the analytical details are in the PRA design records rather than the SAR. Commented [A26]: It is not clear how, for the duration of the operating license, when NRC inspects a licensees conformance NRC is in the process of generating a regulatory guide (DG-1404) that will address the with its 50.59-type requirements, the inspectors will be able to delve into the PRA details relied upon by the licensee to justify a acceptability of using NEI 21-07 to develop portions of an advanced non-LWR SAR. NRC also change not being noticed to NRC in advance.

plans to issue guidance for developing the remaining portions of the SAR (i.e., those portions not covered by NEI 21-07) and for other elements of a license application as part of its Advanced Reactor Content of Application Project (ARCAP).1 Commented [A27]: Although only NEI 21-07 addresses SAR content developed using LMP, it should be noted that the LMP-1.31.4 Application of this Guidance based 50.59 process should be able to address the entire SAR addressed by ARCAP, including those portions not informed by the LMP process.

Sections 2, 3, and 4 of this document provide change control guidance for advanced reactors following NEI 18-04 and NEI 21-07. The guidance is intended to be applied in conjunction with the existing change control guidance in NEI 96-07, with appropriate additions and adjustments as provided herein.

Section 2 of this document addresses the introductory material in Sections 1, 2, and 3 of NEI 96-07.

Section 3 of this document addresses the implementation guidance in NEI 96-07 Section 4. This section covers the three major areas of applicability, screening, and evaluation.

Section 4 of this document addresses documentation and reporting as covered in NEI 96-07 Section 5.

Section 5 of this document provides an overall summary.

1 Slides from the February 25, 2021, NRC Advanced Reactor Stakeholder Meeting provide information on the ARCAP project and its relationship to NEI 21-07. See ML21055A541 pp.91-105.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

2.0 NEI 96-07 INTRODUCTORY MATERIAL

2.1 Introduction (NEI 96-07 Section 1.0)

The information in NEI 96-07 Section 1.0 is applicable to reactors with an LMP-based affirmative safety case licensed under 10 CFR Part 50 or 10 CFR Part 52.

2.2 Defense-in-Depth Design Philosophy and 10 CFR 50.59 (NEI 96-07 Section 2.0)

Section 2.0 of NEI 96-07 discusses the philosophy of DID for LWRs, the role of the General Design Criteria for LWRs documented in 10 CFR 50 Appendix A, and the importance of the UFSAR accident analyses.

The NEI 96-07 Section 2.0 discussion of DID is not directly relevant to an advanced non-LWR with an LMP-based affirmative safety case. There are important differences between the Commented [A28]: Portions could still be relevant to LWR-treatment of DID in 10 CFR 50.59 and NEI 96-07, on the one hand, and the NEI 18-04 definition SMRs using the LMP process.

and evaluation of DID, on the other. The former focuses on the performance of fission product barriers, including fuel, coolant pressure boundary, and containment. The three-barrier LWR DID model is specific to the LWR technology and may not apply to advanced non-LWR designs.

In contrast, NEI 18-04 uses a layers-of-defense concept that addresses plant capabilities, programs, and a risk-informed, performance -based evaluation of DID.

The 10 CFR 50 Appendix A, General Design Criteria are written explicitly for LWRs and are not applicable to advanced non -LWRs. Principal Design Criteria for LWRs are generally derived from 10 CFR 50 Appendix A. In contrast, for advanced non -LWRs, NEI 21-07 SAR Chapter 5 describes a systematic approach for deriving Principal Design Criteria. Therefore, the discussion Commented [A29]: DG-1404 and ARCAP Roadmap ISG also of General Design Criteria in NEI 96-07 Section 2.0 is not applicable to advanced non-LWRs. provide pertinent discussion and clarification PDCs proposed by an applicant using the LMP process.

NEI 96-07 states, The UFSAR presents the set of limiting analyses required by NRC. Commented [A30]: Still applicable to LWR-SMRs though.

Typically, these analyses are deterministic in nature and follow the NRCs Standard Review Plan for LWR accident analyses. In contrast, NEI 18-04 provides for a systematic approach to developing LBEs for advanced non-LWRs.

The fundamental conclusion of NEI 96-07 Section 2 is that:

Changes to plant design and operation and conduct of new tests and experiments have the potential to affect the probability and consequences of accidents, to create new accidents and to impact the integrity of fission product barriers. Therefore, these activities are subject to 10 CFR 50.59.

As discussed above, there are a number of elements of NEI 96-07 Section 2 that are not applicable to advanced non-LWRs. However, the fundamental conclusion of the section holds Commented [A31]: Same comment as previous regarding for non-LWRs, with one caveat. Reactors with an LMP-based affirmative safety case use a safety case.

layers-of-defense approach to safety that is more holistic than the LWR approach, which focuses Commented [A32]: Please review the use of holistic. It isnt a on three fission product barriers to provide DID. From a practical standpoint, this requires an clearly defined regulatory term and may have different meanings.

adjustment to how the 10 CFR 50.59(c)(2)(vii) criterion related to fission product barriers is The idea meant to be conveyed may be better explained in two sentences rather than one?

implemented. The adjustment is addressed in Section 3.3.1. Commented [A33]: Need to keep in mind the LWR-SMRs using LMP.

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2.3 Definitions and Applicability of Terms (NEI 96-07 Section 3.0)

The definitions and applicability criteria presented in NEI 96-07 Section 3.0 are applicable to reactors with an LMP-based affirmative safety case, with the caveats and clarifications provided below. These caveats and clarifications should be applied to all phases of implementation guidance (applicability, screening, and evaluation).

Accident Previously Evaluated in the FSAR (as Updated) (NEI 96-07 Section 3.2)

Reference is made in the definition of accidents, such as those typically analyzed in FSAR Chapters 6 and 15 of the UFSAR. That is appropriate for a currently-operating LWR, but Commented [A34]: May still be appropriate for LWR-SMRs NEI 21-07 provides an alternate organization of material for a reactor with an LMP-based using LMP. Remains to be seen what SAR organizational model an affirmative safety case. The appropriate reference for a reactor following NEI 18-04 would be to LWR-SMR applicant using LMP may choose to use.

SAR Chapters 2 and 3 per NEI 21-07.

The discussion states that the term accidents includes anticipated (or abnormal) operational transients and postulated design basis accidents as well as other events for which the plant is required to cope and that are described in the UFSAR For a reactor with an LMP-based affirmative safety case, the first category of accidents is defined as the LBEs (AOOs, DBEs, Commented [A35]: LMP has transitioned to using event-BDBEs, and DBAs), which, as noted above, are documented in SAR Chapters 2 and 3 per sequences instead of accidents. Consider being more consistent with the NEI 18-04 terminology.

NEI 21-07. The second category remains the same for advanced non-LWRs, to the extent the Commented [A36]: This text indicates that there are events other events are applicable per the regulations. for which the plant is required to cope (second category) that are something beyond the four categories of LBE event sequences. Is Change (NEI 96-07 Section 3.3) this the intent? If so, how is this second category of event sequences established? (Is aircraft impact an example of this second category?)

The definition of change as presented in NEI 96-07 is also applicable to a reactor with an LMP-based affirmative safety case. However, the discussion under the definition in NEI 96-07 addresses the terms design functions and design bases functions. The systematic nature of the NEI 18-04 process allows for a much more straightforward approach to delineating design bases functions and design functions.

For the purpose of evaluating changes to a reactor with an LMP-based affirmative safety case, design bases functions correspond to Required Safety Functions per NEI 21-07 SAR Section 5.2. Design functions are considered to be composed of the design bases functions (Required Safety Functions), risk-significant functions per NEI 21-07 SAR Section 5.5.1, and safety functions required for adequate DID per NEI 21-07 SAR Section 5.5.2.

Malfunction of an SSC Important to Safety (NEI 96-07 Section 3.9)

The definition implies that SSCs important to safety are those with design functions described in the UFSAR (whether or not classified as safety-related in accordance with 10 CFR 50, Appendix B). For the purpose of evaluating changes to a reactor with an LMP-based affirmative safety case, SSCs important to safety is interpreted to be the population of SSCs that are either safety-related or NSRST SSCs, as defined by NEI 18-04. This population of SSCs is also referred to as the safety-significant SSCs. Commented [A37]: This is consistent with how the NRC interpreted SSCs important to safety with respect to PDCs (i.e.,

includes SR and NSRST SSCs).

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Safety Analyses (NEI 96-07 Section 3.12)

The definition of safety analyses notes that containment, emergency core cooling system, and accident analyses in Chapters 6 and 15 of the UFSAR clearly fall within the meaning of safety analyses, recognizing that safety analyses are not limited to those two chapters. Per the discussion in Section 2.3.1 above, those particular types of analyses (if applicable to an advanced reactor with an LMP-based affirmative safety case) would be found in SAR Chapters 2 and/or 3 as defined in NEI 21-07. Commented [A38]: Chapter 4 (integrated analyses) should likely be included since the safety analyses using LMP may also credit NSRST SSCs.

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3.0 IMPLEMENTATION GUIDANCE

The process for 10 CFR 50.59 is shown in Figure 1 of NEI 96-07. This document assumes that a similar process is followed for an advanced non-LWR but that elements of the process are adjusted to reflect the different nature of the reactor and the safety case. The remainder of this section addresses the necessary adjustments to the three facets of the 10 CFR 50.59 process:

applicability, screening, and evaluation. Figure 1 of this document shows the process for an LMP-based affirmative safety case, with the specific evaluation criteria summarized.

NEI 96-07 Section 4 provides implementation guidance for 10 CFR 50.59, and necessary modifications to the guidance are addressed in Section 3.0 of this white paper. The guidance provided in NEI 96-07 Appendix C should also be considered by licensees of nuclear power plants licensed under 10 CFR Part 52. Commented [A39]: This only skims the surface and additional considerations are warranted depending upon the need (i.e.,

perhaps the initial focus is on the near term CPs and OLs under Part 50 rather than DCs using LMP). See previous comments about 50.59-like change process included in rules for DCs and how a 50.59-like change process might look like for an LMP-based DC.

Figure 1. 10 CFR 50.59 Process for a Reactor with an LMP-based Affirmative Safety Case Commented [A40]: This portion of the figure should better align Figure 1 in NEI 96-07

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Figure 1 (Contd)

3.1 Applicability

Section 4.1 of NEI 96-07 addresses applicability. 10 CFR 50.59 is applicable to advanced reactors licensed under 10 CFR Part 50 or 10 CFR Part 52, including those with an LMP-based

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affirmative safety case, for which this additional guidance is provided. Note that NEI 96-07 Appendix C provides additional guidance for new nuclear power plants licensed under 10 CFR Part 52. Commented [A41]: See previous comment in Section 3.0 on Appendix C.

In general, the existing guidance provided in NEI 96-07 Section 4.1 is applicable to advanced reactors with an LMP-based affirmative safety case. However, there may be portions of the guidance that are not applicable due to the characteristics of (i) the reactor or (ii) the reactors safety case. It is not possible to identify in advance all potential instances in which the NEI 96-07 guidance is not appropriate or cannot be applied. However, this section highlights examples where the application of NEI 96-07 Section 4.1 will need to be modified.

Applicability to Licensee Activities (NEI 96-07 Section 4.1.1)

NEI 96-07 makes it clear that certain licensee activities are controlled by other parts of the regulation and are excluded by 10 CFR 50.59(c)(4). This exclusion also applies to a reactor with an LMP-based affirmative safety case. One of the exclusion examples provided in NEI 96-07 is 10 CFR 50.46, the emergency core cooling system regulation. The regulation specifically applies to boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding. It is likely advanced non-LWRs will not use zircaloy or ZIRLO cladding. The NEI 18-04 guidance was written for advanced non-LWRs, so it can be assumed that this example exclusion from 10 CFR 50.59 will not be applicable to reactors with an LMP-based affirmative safety case.

Maintenance Activities (NEI 96-07 Section 4.1.2)

The NEI 96-07 Section 4.1.2 guidance in its entirety is applicable to reactors with an LMP-based affirmative safety case.

UFSAR Modifications (NEI 96-07 Section 4.1.3)

The NEI 96-07 Section 4.1.3 guidance in its entirety is applicable to reactors with an LMP-based affirmative safety case.

Changes to Procedures Governing the Conduct of Operations (NEI 96-07 Section 4.1.4)

The NEI 96-07 Section 4.1.4 guidance in its entirety is applicable to reactors with an LMP-based affirmative safety case.

Changes to Approved Fire Protection Programs (NEI 96-07 Section 4.1.5)

The NEI 96-07 Section 4.1.5 guidance in its entirety is applicable to reactors licensed under 10 CFR Part 50 with an LMP-based affirmative safety case. 10 CFR Part 52 licensees should refer to the guidance provided in NEI 96-07 Appendix C Section 4.1. Commented [A42]: This may not be true for LMP-based applicants. Need to confirm based on the guidance included in Changes to the Probabilistic Risk Assessment (PRA) ARCAP.

PRAs for currently-licensed LWRs are not subject to change control under 10 CFR 50.59. The same approach is retained for advanced non-LWRs that follow NEI 18-04. As described in

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Section 1.2.2, the PRA plays a much more significant role in the LMP-based affirmative safety case than it does in the licensing basis of LWRs licensed under 10 CFR Part 50 and 10 CFR Part 52. The PRA is a living plant model that is kept up to date for many reasons, including to ensure that it adequately represents both the probability and the consequences of the AOO, DBE, and BDBE licensing basis events.

It is neither desirable nor necessary to evaluate changes to the PRA under 10 CFR 50.59.

Instead, this guidance assumes that the licensee has committed to following the Non-LWR PRA Standard,1 which is the controlling document for changes to the PRA. If a licensee does not follow the Non-LWR PRA Standard, then it is incumbent on the licensee to establish with the NRC an acceptable alternative for PRA change control.

Further discussion of the PRA is provided in Appendix A.

Changes to the State of Knowledge

New information relevant to a reactor safety case may be obtained at any time. This is true of currently-licensed LWRs, but these types of changes for new non-LWRs may be more common, at least initially, than for LWRs as a result of refinements in the knowledge of new non-LWRs from operating experience, experiments, and testing. Changes to the state of knowledge are not potential facility changes that are being contemplated; instead, they are actual changes to the best understanding of reality that have already occurred. There is nothing elective about them, and there is nothing to submit to the NRC for approval. Changes to the state of knowledge may impact the regulatory process in other ways, and they may lead to other changes that are subject to a 10 CFR 50.59 screening and potentially evaluation (e.g., a change to a method of evaluation due to an evolution of the understanding of a particular physical phenomenon, a plant modification to regain margin loss due to a change in the state of knowledge, and a plant modification to take advantage of margin gained by a change in the state of knowledge).

However, a change to the state of knowledge, in and of itself, is not subject to a 10 CFR 50.59 review. Commented [A43]: 1)Does this changes to the state of knowledge contemplate a type of operating experience 3.2 Screening program for advanced reactors that include elements such as a generic issues program, etc? 50.54(f)?

2)Since the initial population of non-LWRs of a specific design Section 4.2 of NEI 96-07 addresses screening. Once it has been determined that 10 CFR 50.59 is will be small, the new understanding of reality might occur in an operating plant of a particular design, i.e., no one else has applicable to a proposed activity (see Section 3.1 above), screening is performed to determine if learned of it yet. If there is no obligation under a 50.59-type the activity should be evaluated against the evaluation criteria of 10 CFR 50.59(c)(2). If so, the process to notify the NRC, how else would the NRC learn of the evaluation should be performed as provided for in Section 3.3 below. new information?

The guidance provided in NEI 96-07 Section 4.2 is applicable to advanced reactors with an LMP-based affirmative safety case. With that being said, the documentation of the safety case in a SAR that follows the guidance in NEI 21-07 should enable a relatively straightforward Commented [A44]: Since NEI 21-07 is limited to the first 8 determination of whether or not the criteria associated with screening a change in (50.59 chapters of the SAR, it would be helpful to include some discussion that the 50.59 process applies to the entire SAR not just those evaluation required) are satisfied. If not, the activity screens out, and a 50.59 evaluation is not portions informed by NEI 21-07 (see ARCAP Roadmap ISG). Dont want to create the impression that 50.59 is just limited to the first 8 chapters of an LMP-based SAR.

1 ANSI/ASME/ANS RA-S-1.4-2021, Probabilistic Risk Assessment Standard for Advanced Non-Light Water Reactor Nuclear Power Plants, American Society of Mechanical Engineers and American Nuclear Society, approved January 28, 2021.

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needed. This section highlights those aspects of the LMP-based affirmative safety case and associated SAR documentation that are particularly pertinent to screening.

Change to the Facility or Procedures (NEI 96-07 Section 4.2.1)

In screening, an essential step is to determine whether or not a proposed activity affects a design function, method of performing or controlling a design function or an evaluation that demonstrates that design functions will be accomplished For the LMP-based affirmative safety case, design functions should be documented in Chapter 5 of the SAR per NEI 21-07: a Required Safety Function in Section 5.2, a risk-significant function in Section 5.5.1, and a safety function required for adequate DID in Section 5.5.2. See Section 2.3.2 of this document for additional discussion.

Changes to the Facility as Described in the UFSAR (NEI 96-07 Section 4.2.1.1)

SSCs that are relied upon to carry out design functions are documented in the NEI 21-07 SAR in Section 5.4 (safety-related or SR) and Section 5.5 (non-safety-related with special treatment or NSRST). However, as addressed in Section 4.2.1.1, changes to other SSCs (i.e., no special treatment or NST) should be considered for potential adverse effects on any SR or NSRST SSC design function, method of performing or controlling the design function, or an evaluation demonstrating that the intended design functions will be accomplished.

In accordance with NEI 18-04, reliability and capability targets are documented in NEI 21-07 SAR Sections 6.2 and 7.1 for SR and NSRST SSCs, respectively. If a proposed change to the facility results in a safety-significant SSC being unable to meet its reliability or capability target, then the change would screen in, and a full 10 CFR 50.59 evaluation would be required. Commented [A45]: Wouldnt this require the reliability and capability targets to be included in the SAR or incorporated by Changes to Procedures as Described in the UFSAR (NEI 96-07 Section 4.2.1.2) reference in the SAR?

Procedures should be screened in only if they affect design functions (see Sections 2.3.2 and Commented [A46]: Could this also include those situations 3.2.1 above). Required operator actions should be addressed in the SAR documentation of the where an alternative means of performing the credited design associated SSCs, provided in NEI 21-07 SAR Chapter 6 (SR SSCs) and NEI 21-07 SAR functions is proposed than the manner credited for providing the design function in the SAR?

Chapter 7 (NSRST SSCs).

In an analogous manner to facility changes discussed in Section 3.2.2 above, if a proposed change to a procedure results in a safety-significant SSC being unable to meet its reliability or capability target, then the change would screen in and a full 10 CFR 50.59 evaluation would be required. Commented [A47]: See above comment on including reliability and capability targets in the SAR or incorporated by reference in the Changes to UFSAR Methods of Evaluation (NEI 96-07 Section 4.2.1.3) SAR.

Methods of evaluation associated with DBAs should be addressed in NEI 21-07 SAR Sections 2.2 (Source Term), 2.3 (DBA Analytical Methods), and 3.6 (Design Basis Analyses). Commented [A48]: What is the basis for limiting to DBA Adverse changes to DBA methods would screen in. Methods of evaluation associated with the analyses? Couldnt other analyses potentially affected by changes in methods of evaluation have a potential adverse impact on DBA remaining LBEs (AOOs, DBEs, and BDBEs) should be addressed in the PRA and are therefore results? Margins of safety as defined in TS? Ability of SSCs to not applicable to further screening or evaluation (see Section 3.1.6 above). Methods of meeting reliability or capability targets? Etc evaluation not associated with LBEs may be addressed in NEI 21-07 SAR Section 2.4 (Other Commented [A49]: What is an adverse change? There is no clear definition to aid in consistent application.

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Methodologies and Analyses) or in other parts of the SAR not covered by NEI 21-07 guidance.

Adverse changes to non-LBE methods of evaluation would screen in.

Test or Experiment Not Described in the UFSAR (NEI 96-07 Section 4.2.2) Commented [A50]: What about significant operating events not evaluated in the SAR (e.g., component or human failures)?

Tests or experiments described in the SAR may be located in NEI 21-07 SAR Chapter 2, NEI 21-07 SAR Section 6.3 (SR SSCs), and NEI 21-07 SAR Section 7.2 (NSRST SSCs). If already described in the SAR, the tests or experiments would screen out. Commented [A51]: What about testing described in programs discussed in SAR sections not discussed in NEI 21-07?

3.3 Evaluation

If a planned change has reached the evaluation portion of the 10 CFR 50.59 process, an applicability evaluation has determined that 10 CFR 50.59 is applicable to the proposed activity (see Section 3.1). In addition, screening has determined the activity is (i) a test or experiment not described in the UFSAR or (ii) a modification, addition, or removal (i.e., change) that adversely affects a design function of an SSC, a method of performing or controlling the design function, or an evaluation for demonstrating that intended design functions will be accomplished (see Section 3.2). At this point, the licensee would perform a detailed evaluation of the adverse effect of the activity against the eight criteria of 10 CFR 50.59(c)(2).

It is in the evaluation portion that the most significant changes arise relative to the existing NEI 96-07 guidance for light water reactors with a traditional deterministic safety case. The risk-informed, performance-based approach to establishing an LMP-based affirmative safety case is very conducive to the 10 CFR 50.59 evaluation. Elements of the LMP-based affirmative safety case, such as risk significance and DID, enable an objective evaluation against the eight criteria, as described in this section. However, the evaluation process is modified somewhat from the approach of the existing NEI 96-07 guidance, as described in the remainder of this section. The need to modify the process stems from the differences in terminology and substance between an LMP-based affirmative safety case and a traditional deterministic safety case.

Evaluation Criteria

As shown in Table 1, evaluation criteria derived from NEI 18-04 have been established that enable the licensee to determine whether or not the intent of the evaluation criteria in 10 CFR 50.59(c)(2) are met. These alternative criteria based on NEI 18-04 are necessary to account for the risk-informed and performance-based nature of an LMP-based affirmative safety case. The eight evaluation criteria listed in 10 CFR 50.59(c)(2) are listed in the first column of Table 1. These criteria have been reordered and grouped into three categories to put them into the context of an LMP-based affirmative safety case. The second column of Table 1 provides the functionally equivalent criteria for a licensee following NEI 18-04. These criteria are referred to as LMP 50.59 criteria. The third column of Table 1 provides explanatory comments.

The first category of criteria covers changes that impact the frequency or consequences of accidents [criteria (i), (iii), and (v)]. The second category addresses changes that impact SSCs

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[criteria (ii), (iv), (vi), and (vii)]. The third category consists of criterion (viii) - changes to evaluation methods.

The LMP 50.59 criteria in Table 1 refer to LBEs, which are composed of AOOs, DBEs, BDBEs, and DBAs. The term accident is not used in NEI 18-04 or the Non-LWR PRA standard.

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Table 1. 10 CFR 50.59 Evaluation Criteria for an LMP-based Affirmative Safety Case LMP 50.59 Criteria for an LMP-based Commented [A52]: 53.1550 is the proposed Part 53 change-10 CFR 50.59(c)(2) Criteria Affirmative Safety Case Comments process control regulation (similar to 50.59). The proposal in 53.1550 is similar to this except includes criterion (2)(ii) on margin Category 1 - Accidents reduction for risk-significant LBEs Commented [A53]: Regarding the handling of uses of the LMP (i) Result in more than a minimal increase (a) Result in a change to the frequency or Risk significance of an LBE in the LMP involving design targets such as maintaining doses below 1 rem at in the frequency of occurrence of an consequences of one or more AOOs, context and in the Non-LWR PRA the EAB for consideration of alternative EPZs, would this criterion be revised to include the design target as well as the F-C target?

accident previously evaluated in the final DBEs, or BDBEs documented in the final standard requires the consideration of Commented [A54]: The facilitys FSAR reflects the frequency safety analysis report (as updated); safety analysis report (as updated) in a the combination of frequency and and consequence of the identified LBEs, and NRCs SER accepts manner that would exceed the NEI 18-04 consequence effects. There are no them based on the margins to the associated requirement(s). For (iii) Result in more than a minimal Frequency-Consequence Target or criteria to evaluate these components of instance, for an alternate EAB, NRC may assess the LBE margin to an offsite consequence of 1 rem. So, it seems that pending changes increase in the consequences of an change an LBE from non-risk significant to risk separately. should be assessed against how they impact this original licensing accident previously evaluated in the final basis margin, rather than on the high level targets (like the F-C risk significant according to NEI 18-04 LBE targets) which are likely orders of magnitude higher in consequence safety analysis report (as updated); risk significance criteria. and well beyond the consequences accepted by NRC in the facilitys LMP DBAs are evaluated SER..

[See NEI 96-97 Sections 4.3.1 and 4.3.3, (b) Result in more than a minimal deterministically, like LWR accidents. Commented [A56]: This is a very important distinction from existing 50.59 evaluation criteria where just a change in frequency respectively] increase in the consequence of a Design Therefore, determining if a change leads (i.e., an increase) could result in an unreviewed safety question.

Basis Accident documented in the final to a more than minimal increase in DBA Now the criteria is a change in frequency and consequence that safety analysis report (as updated). consequences should follow the existing moves it above F-C curve.

NEI 96-07 Section 4.3.3 guidance. Commented [A57]: See previous comment regarding options applicants have in determining source term. Mechanistic source term may be used as part of DBA and, therefore, is not like LWR v) Create a possibility for an accident of a (c) Result in one or more AOO, DBE, or Newly identified LBEs or changes to LBE accidents different type than any previously BDBE that is (i) not previously evaluated frequencies and consequences that are Commented [A55]: Note similar to 53.1550 (new criterion (2)(vii) while criterion (2)(vi) addresses create new DBA evaluated in the final safety analysis in the UFSAR and (ii) classified as risk not risk significant should be documented Commented [A58]: Note that 53.1550 does not include a report (as updated); significant according to NEI 18-04 LBE risk in the next final safety analysis report criterion for new AOO, DBE, BDBE but would be possible to tweak significance criteria. update, but the associated change does (2)(1) to include new LBE as well as increase f/c of existing LBE

[See NEI 96-07 Section 4.3.5] not require prior NRC review. Commented [A59]: Note that 53.1550 includes a criterion for

. cumulative risk measures - discuss extent to which this is addressed in the DID or other criterion in this table

Note that as mentioned in some other comments, 53.1550 Criterion (2)(v) addresses margin reductions related to design goals such as 1 rem for EPZ. This warrants a good discussion since many applications expected to propose such design goals for EPZ, siting, staffing, etc.

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LMP 50.59 Criteria for an LMP-based 10 CFR 50.59(c)(2) Criteria Affirmative Safety Case Comments Category 2 - SSCs (ii) Result in more than a minimal (d) Result in an increase in the frequency 10 CFR 50.59(c)(2) criteria (ii), (iv), (vi), Commented [A60]: Note that 53.1550 does not have a similar increase in the likelihood of occurrence of or consequences of a malfunction of any and (vii) are addressed collectively by criterion on the risk-significance on an SSC. This is a good addition to capture the effect of a change on specific SSCs.

a malfunction of a structure, system, or safety-significant SSC that would change LMP 50.59 criteria (d) and (e).

component (SSC) important to safety the classification of the SSC from non -risk previously evaluated in the final safety significant to risk-significant. Changes with the impacts on the LMP-Commented [A61]: How does this result in a similar analysis report (as updated); based affirmative safety case described in effectiveness to 96-07 section 4.3.2?

(e) Result in an increase in the frequency (d) or (e) are deemed to require prior Commented [A62]: Note that 53.1550 does not have a similar (iv) Result in more than a minimal or consequences of a malfunction of a NRC approval. criterion on DID. This is a good addition but perhaps try to avoid more than minimal since DID is difficult to quantify. NST to NSRST increase in the consequences of a safety-significant SSC that would have a is good but would something else be needed to capture different malfunction of an SSC important to safety more than minimal adverse effect on role for existing NSRST SSC in helping ensure DID.

previously evaluated in the final safety defense-in-depth adequacy or lead to a Commented [A63]: Possible question similar to above - this analysis report (as updated); change in safety classification from NST criterion could be more complex than initially seems. Most changes are not actually revising the overall plant configuration and so it to NSRST to maintain adequate defense - would be simple to say no impact on DID, but more significant plant

[See NEI 96-07 Sections 4.3.2 and 4.3.4, in-depth. changes might warrant reassessment by an IDPP.

respectively] Commented [A64]: How is more than minimal adverse effect on DID adequacy defined?

(vi) Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the final safety analysis report (as updated);

[See NEI 96-07 Section 4.3.6]

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LMP 50.59 Criteria for an LMP-based 10 CFR 50.59(c)(2) Criteria Affirmative Safety Case Comments (vii) Result in a design basis limit for a No specific criterion The DID provided by LWR fission product fission product barrier as described in the barriers is addressed in a holistic manner FSAR (as updated) being exceeded or in NEI 18-04. There is no need to single altered; out fission product barriers in LMP 50.59 criteria; instead, impacts of changes on all

[See NEI 96-07 Section 4.3.7] safety-significant SSCs (not just fission product barriers) and DID are addressed by LMP 50.59 criteria (d) and (e).

Category 3 - Methods of Evaluation (viii) Result in a departure from a method (f) Result in a departure from a method of Evaluation of changes to methods of Commented [A65]: Note that this is a bit different from of evaluation described in the FSAR (as evaluation described in the FSAR (as evaluation should follow NEI 96-07 53.1550 with a focus on the DBA and coverage of other LBEs via the PRA standard. Criterion (2)(iv) in 53.1550 does allow credit for NRC updated) used in establishing the design updated) used in establishing the design Section 4.3.8 guidance. approved standards which might align with this table. Might tweak bases or in the safety analyses. bases or in the safety analyses, with the 53.1550 to capture DBAs.

exception of LBE evaluation methods Note that methods of evaluation used in Commented [A66]: This would not limit this criteria to just the

[See NEI 96-07 Section 4.3.8] under the change control of the Non-LWR the PRA are not addressed by 10 CFR results of the DBA analysis.

PRA Standard. 50.59 (see Section 3.1.6 of this guidance). Commented [A67]: May want to reword this since "design bases" has mixed meanings - the design bases of specific SSCs (50.2)

Such methods are instead managed by or more general use of the term?

New Methods and Configuration Control requirements in the Non-LWR PRA Standard ASME/ANS RA-S-1.4-2021.

These include methods of evaluation for AOOs, DBEs, and BDBEs.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

Evaluation Process

The process for evaluating a potential change to a facility to determine the need for prior NRC approval is described in this section. The process addresses the LMP 50.59 criteria listed in the center column of Table 1. If none of the criteria are satisfied, then the change may be implemented without prior NRC approval. If any LMP 50.59 criterion is satisfied, then the proposed activity requires prior NRC approval. The two potential outcomes of the evaluation are described further in NEI 96-07 Section 4.5.

The licensee must address all LMP 50.59 criteria (the entire center column of Table 1) in order to substantiate that a conclusion that prior NRC approval is not required. The evaluation can be performed in any order, but this guidance assumes the licensee will step through the criteria in Categories 1, 2, and 3, respectively, as provided in Section 3.3.1. Thus, the licensee would first consider LBE impacts, then SSC impacts, and finally methods of evaluation impacts. Further guidance for the application of the criteria is provided below.

Category 1 - Accidents 10 CFR 50.59(c)(2)(i) and (iii) refer to the frequencies and consequences, respectively, of accidents. Under current NEI 96-07 guidance accidents include AOOs, DBAs, and other events added to the licensing basis and reflected in the SAR such as anticipated transients without scram and station blackout for light water reactors. NEI 18-04 systematically identifies a range of LBEs that include AOOs, DBEs, BDBEs, and DBAs. In order to meet the requirements of the Non-LWR PRA Standard, the AOOs, DBEs, and BDBEs involve all credible combinations of safety system successes and failures, including consideration of common cause failures of the type involved in anticipated transient without scram and station blackout-type sequences. They also include, as appliable, event sequences involving multiple reactors and radionuclide sources. For the purpose of evaluating criteria (i) and (iii) in Category 1, this guidance addresses all four types of NEI 18-04 LBEs.

Information on LBE frequency and consequence is developed in the PRA and is provided in the SAR for AOOs, DBEs, and BDBEs. DBAs are defined using a set of deterministic rules that involve the selection of safety-related SSCs in the performance of Required Safety Functions and the subsequent evaluation of their consequences.are evaluated deterministically under specific rules - the DBAs have no associated frequency. For the purpose of evaluating LBEs against the Commented [A68]: But are derived from the DBEs which do Category 1 LMP 50.59 criteria, AOOs, DBEs, and BDBEs are addressed in terms of frequency, have a frequency range associated with them (i.e., 10E-2 to 10E-4) consequence, and risk, whereas DBAs are addressed deterministically. Commented [A69]: This is not exactly true. DBAs do have a frequency, although its not used in the event selection and evaluation process. So - it would be better to delete this text and In the NEI 18-04 methodology, frequencies and consequences of AOOs, DBEs, and BDBEs are modify the sentence with something like the markup indicated.

not evaluated separately but rather against risk criteria that consider the combination of frequency and consequences. As shown on NEI 18-04 Figure 3-1, AOOs, DBEs, and BDBEs are expected to be to the left of the Frequency-Consequence (F-C) Target, and are classified as either risk significant or non-risk-significant. For AOOs, DBEs, and BDBEs the term more than minimal increase is interpreted in the context of the risk significance criteria and the F-C Target. These criteria are clearly stated, performance-based, and unambiguous to apply.

For application of LMP 50.59 criterion (a) in Table 1, the changes are deemed to have more than a minimal increase in frequency or consequences if an existing AOO, DBE, or BDBE

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

changes its risk classification from non-risk-significant to risk-significant or if the change increases the risk such that the AOO, DBE, or BDBE exceeds the frequency-consequence (F-C) target (see NEI 18-04 Section 3.2.2, Tasks 7a and 7c). NEI 18-04 Figure 3-4 provides a graphical representation of the risk significant region and the F-C target. The evaluation of LMP 50.59 criterion (a) may be performed by using the PRA to evaluate the effect on risk significance and the F-C target consistent with NEI 18-04. It is noted that a proposed change may also result in changing one or more LBEs from risk significant to non-risk significant. Such a result would not satisfy LMP 50.59 criterion (a) and would therefore not translate to a requirement for prior NRC review of the proposed change. Commented [A70]: With the way this reads, it appears that that an LBE going from a DBE (which DBAs are derived from) to a LMP 50.59 criterion (b) impacts NEI 18-04 DBAs only. Such DBAs are analyzed in a manner BDBEE may result in an SSC going from being in Tech specs to not being in tech specs. This issue should be acknowledged in this consistent with LWR Chapter 15 events, i.e., they are evaluated deterministically against the document through an additional sentence, possibly like: While LMP same consequence criteria as in the current regulations, 10 CFR 50.34 and 10 CFR 100. 50.59 criterion (a) may not be met in such a case, other criterion may apply that would require prior NRC approval (e.g., if a technical Therefore, the language in criterion 50.59(c)(2)(iii) is applicable to DBAs, and it has been specification is no longer required based on changing an LBE from retained in LMP 50.59 criterion (b). If the proposed change affects the plant response to an NEI risk-significant to non-risk significant).

18-04 DBA, the effect of the change on the DBA should be assessed consistent with the Commented [A71]: See previous comment about optional use guidance in NEI 96-07 Section 4.3.3 to determine if there is a more than a minimal increase in of mechanistic source term even for DBA analysis.

the consequences. Note that criterion (b) addresses only consequences and not frequencies of DBAs. Because DBAs are derived from DBEs, LMP criterion (a) implicitly addresses the frequency of the underlying LBE, and no additional treatment is necessary for DBA frequency. Commented [A72]: As a result, the F-C target curve does not come into play for this criterion. Consistent with 96-07, the Changes that may introduce newly identified LBEs are addressed by LMP 50.59 criterion (c). comparison is to the analyzed DBA dose.

Such changes should be evaluated by revisiting the NEI 18-04 process for identifying LBEs after quantifying the risk significance of the new LBE using the PRA. The evaluation of the change should determine whether there are new initiating events introduced by the change or whether the change alters the event sequence plant response model in a manner that introduces a new event sequence or event sequence family. It is important to note that criterion (c) is satisfied only when any new LBEs exceed the risk significance criteria in NEI 18-04 based on mean values of frequency and consequence. If a newly identified LBE is not risk significant, it has no Commented [A73]: Why isnt the determination based on 95%

material impact on the LMP-based affirmative safety case, so the change would not require prior confidence plus or minus 5% to account for cliff-edge effects?

Seems like the evaluation here based on mean values is of lesser NRC review. Also, because a new DBA would require a new DBE, LMP 50.59 criterion (c) veracity than the original.

covers DBAs as well as AOOs, DBEs, and BDBEs.

Category 2 - SSCs The next category of evaluation criteria addresses changes that impact the performance of SSCs identified in the UFSAR. 10CFR50.59(c)(2) includes several criteria associated with SSCs that are deemed important to safety (ITS). As discussed in Section 2.3.3 of this document, NEI 18-04 does not use important to safety but rather uses two SSC categories that collectively are regarded as safety significant: SR and NSRST. NSRST SSCs are so classified when the SSC functions either meet SSC risk significance criteria or provide functions that are deemed necessary for adequate DID. The SSC criteria of this change control guidance address safety -

significant SSCs, and in doing so, address DID adequacy as well.

Criteria 10 CFR 50.59(c)(2)(ii), (iv), and (vi) address the likelihood and consequence of SSC malfunctions separately. NEI 18-04 SSC risk significance criteria define SSC risk significance based on a combination of frequency or probability of occurrence and consequences of failure.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

Note that this is consistent with the holistic treatment of LBE risk significance which was addressed above under Category 1. Therefore, the 10 CFR 50.59 criteria for SSCs that correspond to more than minimal increase in the likelihood of failure or consequences of an SSC malfunction are evaluated using LMP 50.59 criterion (d) (based on NEI 18-04 SSC risk significance) and LMP 50.59 criterion (f) (based on NEI 18-04 SSC safety classification).1

Changes that may impact SSC risk significance or safety classification should be evaluated by revisiting the pertinent processes in NEI 18-04 Section 3.2.2 after quantifying the impact on risk using the PRA.

LMP 50.59 criterion (e) addresses adverse effects on DID adequacy. DID is addressed in Chapter 5 of NEI 18-04 and Chapter 4 of NEI 21-07. The focus of the evaluation of the effect on DID adequacy is on the integrated DID evaluation as documented in NEI 21-07 SAR Section 4.2.3, which addresses the adequacy of plant capability and programmatic DID. NEI 21-07 SAR Section 4.2.3 addresses actions to establish DID adequacy described in NEI 18-04 Section 5.9.3.

Note that one of the plant capability DID adequacy criteria is Risk margins against the F-C target are sufficient. These margins are addressed by LMP 50.59 criterion (a) with respect to LBEs changing from not risk significant to risk significant and, therefore, do not require re-evaluation under this criterion (e). However, cumulative risk targets documented in NEI 21-07 Section 4.1 are included in this DID assessment. LMP 50.59 criterion (e) would not be met if the cumulative risk target were to exceed the targets documented in NEI 21-07 Section 4.1. Commented [A74]: Criterion (e) seems not clearly linked to the cumulative risk target consideration. These two sentences do not Any changes which result in a more than minimal adverse effect on DID adequacy would require fully explain the relationship between the DID assessment (i.e.,

Criterion (e)) and the cumulative risk target).

prior NRC approval. The meaning of more than minimal necessarily varies with the design of the plant, the nature of the safety case, and each facet of DID being evaluated. It is intended that the license applicant would, where feasible, establish guidelines up front in NEI 21-07 SAR Section 4.2.3 and the design records, to assist in performing the LMP 50.59 criterion (e) evaluation (i.e., assist in answering the question of whether a change has a more than minimal effect on each facet of integrated DID). Commented [A75]: Is there a possibility that the SAR itself could define more than minimal adverse impact to DID adequacy The nature of the change and its impact on the LMP-based affirmative safety case will impact the or possibly defined in conjunction with a change process included in a license condition?

approach taken to carrying out the DID portion of the 10 CFR 50.59 evaluation. It is anticipated that many changes will be simple and limited in scope such that the evaluation against the LMP 50.59 criteria will be relatively straightforward, using the information and criteria documented in the SAR and the plant records. However, some changes may require a more comprehensive Integrated Decision-Making Process review of DID, including the possibility of utilizing an Integrated Decision-Making Panel, as described in NEI 18-04 Chapters 4 and 5. Once the LMP 50.59 criterion (e) determination is made, the basis for the determination must be documented as discussed in Section 5 of NEI 96-07. If necessary, there should be an update of the DID baseline evaluation in the SAR and plant records.

Note that this approach requires upfront consideration of change control when the DID baseline is established and documented in the SAR and plant records. It may be appropriate to enhance

1 It is noted that the concept of linking more than minimal increase to risk significance thresholds is being used in some of the draft language of 10 CFR Part 53 that addresses change control.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

the guidance in NEI 21-07 to clarify this expectation and ensure it will be accomplished during the establishment of the initial licensing basis. Commented [A76]: It appears that a technology-inclusive guidance such as this paper (or NEI 21-07) should be able to provide There is no LMP 50.59 criterion explicitly addressing 10 CFR 50.59(c)(2)(vii), which focuses on the generic criteria and associated guidelines. Leaving it up to individual SAR could impact consistency as well as inefficiency design basis limits for fission product barriers. There are important differences between the during the licensing review itself.

treatment of DID in 10 CFR 50.59 and NEI 96-07, on the one hand, and the NEI 18-04 definition and evaluation of DID, on the other. The former focuses on the performance of fission product barriers, including fuel, coolant pressure boundary, and containment. The three-barrier LWR DID model is specific to the LWR technology and may not apply to advanced non-LWR designs.

In contrast, NEI 18-04 uses a layers-of-defense concept that addresses plant cap abilities, programs, and a risk-informed, performance -based evaluation of DID. Accordingly, in NEI 18-04, fission product barriers are addressed as part of the safety classification and performance requirements included in the SAR for SSCs in general. Although a traditional LWR fission product barrier may be classified as SR or NSRST under NEI 18-04, its treatment in the LMP-based affirmative safety case is not elevated above other types of SSCs. LMP 50.59 criteria (d), and (e) address SSCs in a comprehensive manner, including fission product barriers to the extent they are applicable, so there is no need to include an explicit fission product barrier LMP 50.59 criterion. Commented [A77]: This discussion could justify being able to use the same LMP 50.59 process for LWR-SMRs that used LMP.

Category 3 - Evaluation Methods For an LMP-based affirmative safety case, LMP 50.59 criterion (f) for evaluation methods is consistent with the 10 CFR 50.59(c)(2)(viii). Note that changes in evaluation methods used in the PRA, including those used for AOO, DBEs, and BDBEs, do not require prior NRC approval because they are addressed through adherence to ASME/ANS RA-S-1.4-2021 (see Section 3.1.6 above).

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

4.0 DOCUMENTATION AND REPORTING (NEI 96-07 SECTION 5.0)

Licensees using the guidance for an LMP-based affirmative safety case should follow the documentation and reporting guidance in Section 5 of NEI 96-07. In documenting the evaluation, the licensee should use the criteria shown in the middle column of Table 1 rather than the standard criteria as worded in 10 CFR 50.59(c)(2). As discussed in Section 3 above, for a licensee following NEI 18-04, the LMP-based criteria are functionally equivalent to the criteria provided in the regulations.

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TIRICE for Non-Light Water Reactors Change Control Scope and Process for a Reactor Licensed in Accordance with the NEI 18-04 Guidance

5.0

SUMMARY

An effective change control program is necessary to ensure that a nuclear power reactor will operate in a safe and efficient manner. This document addresses the application of 10 CFR 50.59, the NRCs requirements for prior approval of facility changes, to advanced non-LWRs with an LMP-based affirmative safety case. The document addresses the three key aspects of the current guidance for LWR change control as discussed in NEI 96-07: applicability, screening, and evaluation. To the extent possible, this guidance takes advantage of the risk-informed, performance-based attributes of reactors which follow the methodologies and guidance provided by NEI 18-04 and NEI 21-07.

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Change Control Scope and Process for a Reactor Licensed Appendix A in Accordance with the NEI 18-04 Guidance

Probabilistic Risk Assessment

The PRA is a representation of important elements of the nuclear power plant facility, and it plays a key role in the NEI 18-04 methodology. The integrated PRA model is actually hundreds of separate models, including system models, event tree models, top logic models, data, etc.,

supporting each of the hazard models as applied to each of the analyzed plant operating states.

Among the many elements of the PRA are the AOO, DBE, and BDBE probability and consequence analyses that comprise a subset of the LBEs for the plant.

The PRA can change due t o periodic updates or as a result of changes in knowledge about the various models. For example, industry data on the reliability of a component may evolve as additional data on it is gathered, and such a change could impact the probability of an event sequence that is one of the plants AOOs, DBEs, and BDBEs. Another example would be a change to the consequence model associated with an AOO, DBE, or BDBE. Yet another would be a decision to model a particular aspect of the plant in additional detail instead of relying on simplified assumptions. It is important that the operator keep the PRA up -to-date, which means modifying it to reflect significant new information and incorporating accurate and reliable models of plant performance.

For the purposes of this guidance, it is assumed that the licensee has committed to follow the Non-LWR PRA standard.1 The standard provides comprehensive guidance for maintaining and updating the PRA. The scope of the information in the PRA makes it both impractical and undesirable for the PRA to be under 10 CFR 50.59 change control. Instead, licensees should be required to follow the Non-LWR PRA standard when updating the PRA, and those activities will be subject to NRC audit and inspection.

With this approach, changes to methods of analyses for AOOs, DBEs, and BDBEs will not be addressed by 10 CFR 50.59. However, such changes will be addressed by a comprehensive and industry-accepted program - the Non-LWR PRA standard - which is expected to be endorsed by an NRC regulatory guide. It should also be noted that DBAs will be analyzed with a traditional Commented [A78]: Trial RG 1.247 has been issued.

conservative Chapter 15 approach to show conformance to dose limits, and those methods of analyses will be subject to 10 CFR 50.59 change control. All in all, non-LWRs that conform to Commented [A79]: See precious comment about option to use NEI 18-04 will have a more systematic and comprehensive safety case than is provided by the mechanistic source term even as part of DBA analysis.

traditional LWR approach.

The PRA tool is a fundamental part of the LMP-based affirmative safety case. In fact, as discussed elsewhere in this white paper, a PRA evaluation of potential plant changes will be a key factor in determining if prior NRC approval for the change is required. PRA results are provided in various sections of a SAR that follows NEI 21-07. If such PRA results change, they will be reflected in the periodic SAR updates.

1 ANSI/ASME/ANS RA-S-1.4-2021, Probabilistic Risk Assessment Standard for Advanced Non-Light Water Reactor Nuclear Power Plants, American Society of Mechanical Engineers and American Nuclear Society, approved January 28, 2021.

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Change Control Scope and Process for a Reactor Licensed Appendix B in Accordance with the NEI 18-04 Guidance

Terminology and Definitions

NEI 18-04 and NEI 21-07 use terminology and definitions specific to reactors approved for operation based on an LMP-based affirmative safety case.

Table B-1 provides the definitions of key terms from the aforementioned documents.

Table B-2 describes how some terms from NEI 96-07 are applied in change control for a reactor following the NEI 18-04 and NEI 21-07 methodology.

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Change Control Scope and Process for a Reactor Licensed Appendix B in Accordance with the NEI 18-04 Guidance

Table B-1. Terminology and Definitions Term Definition Source Anticipated Operational Anticipated event sequences expected to occur one or more times during the life of NEI 18-04 and NEI 21-07 Occurrence (AOO) a nuclear power plant, which may include one or more reactors. Event sequences with mean frequencies of 1x10-2/plant-year and greater are classified as AOOs. AOOs take into account the expected response of all SSCs within the plant, regardless of safety classification.

Beyond Design Basis Event Rare event sequences that are not expected to occur in the life of a nuclear power NEI 18-04 and NEI 21-07 (BDBE) plant, which may include one or more reactors, but are less likely than a DBE. Event sequences with mean frequencies of 5x10-7/plant-year to 1x10-4/plant-year are classified as BDBEs. BDBEs take into account the expected response of all SSCs within the plant regardless of safety classification.

Complementary Design Design criteria for NSRST SSC that are necessary to satisfy the PRA Safety Function(s) NEI 21-07 Criteria (CDC) associated with the SSC. The CDC may be defined at a functional level, or more specifically addressed to the NSRST SSC specific function(s). The CDC for the NSRST SSC are directly tied to the success criteria established in the PRA for the PRA Safety Function(s) responsible for the classification of the SSC as NSRST.

Defense-in-Depth (DID) An approach to designing and operating nuclear facilities that prevents and mitigates NRC Glossary accidents that release radiation or hazardous materials. The key is creating multiple independent and redundant layers of defense to compensate for potential human and mechanical failures so that no single layer, no matter how robust, is exclusively relied upon. Defense-in-depth includes the use of access controls, physical barriers, redundant and diverse key safety functions, and emergency response measures.

Design Basis Accident Postulated accidents that are used to set design criteria and performance objectives NEI 18-04 and NEI 21-07 (DBA) for the design of SR SSCs. DBAs are derived from DBEs based on the capabilities and Commented [A80]: Add definition of DBE along with frequency reliabilities of SR SSCs needed to mitigate and prevent accidents, respectively. DBAs range.

are derived from the DBEs by prescriptively assuming that only SR SSCs classified are available to mitigate postulated accident consequences to within the 10 CFR 50.34 dose limits.

Design Basis Hazard Level A design specification of the level of severity or intensity of a hazard for which the SR NEI 21-07 but (DBHL) SSCs are designed to withstand with no adverse impact on their capability to corrected by removing perform their Required Safety Functions. external.

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Change Control Scope and Process for a Reactor Licensed Appendix B in Accordance with the NEI 18-04 Guidance

Term Definition Source Frequency-Consequence A target line on a frequency-consequence chart that is used to evaluate the risk NEI 18-04 and NEI 21-07 Target (F-C Target) significance of LBEs and to evaluate risk margins that contribute to evidence of adequate Defense-in-Depth.

Licensing Basis Event (LBE) The entire collection of event sequences considered in the design and licensing basis NEI 18-04 and NEI 21-07 of the plant, which may include one or more reactors. LBEs include AOOs, DBEs, BDBEs, and DBAs.

Non-Safety-Related with Non-safety-related SSCs that perform risk-significant functions or perform functions NEI 18-04 and NEI 21-07 Special Treatment that are necessary for Defense-in-Depth adequacy.

SSCs (NSRST SSCs)

PRA Safety Function (PSF) Reactor design specific SSC functions modeled in a PRA that serve to prevent and/or NEI 18-04 and NEI 21-07 Commented [A81]: Add Principal Design Criteria.

mitigate a release of radioactive material or to protect one or more barriers to Add QHO.

release. In ASME/ANS-Ra-S-1.4-2013 these are referred to as safety functions.

The modifier PRA is used in NEI 18-04 to avoid confusion with safety functions performed by SR SSCs.

Required Functional Reactor design-specific functional criteria that are necessary and sufficient to meet NEI 18-04 and NEI 21-07 Design Criteria (RFDC) the Required Safety Functions.

Required Safety Function A PRA Safety Function that is required to be fulfilled to maintain the consequence of NEI 18-04 and NEI 21-07 one or more DBEs or the frequency of one or more high -consequence BDBEs inside the F-C Target.

Risk-Significant LBE An LBE whose frequency and consequence meet a specified risk significance NEI 18-04 and NEI 21-07 criterion. In the LMP framework, an AOO, DBE, or BDBE is regarded as risk-significant if the combination of the upper bound (95th percentile) estimates of the frequency and consequence of the LBE are within 1% of the F -C Target AND the upper bound 30-day TEDE dose at the EAB exceeds 2.5 mrem.

Risk-Significant SSC An SSC that meets defined risk significance criteria. In the LMP framework, an SSC is NEI 18-04 and NEI 21-07 regarded as risk-significant if its PRA Safety Function is: a) required to keep one or more LBEs inside the F-C Target based on mean frequencies and consequences; or b) if the total frequency LBEs that involve failure of the SSC PRA Safety Function contributes at least 1% to any of the LMP cumulative risk targets. The LMP cumulative risk targets include: (i) maintaining the frequency of exceeding 100 mrem to less than 1/plant-year; (ii) meeting the NRC safety goal QHO for individual risk of early fatality; and (iii) meeting the NRC safety goal QHO for individual risk of latent cancer fatality.

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Change Control Scope and Process for a Reactor Licensed Appendix B in Accordance with the NEI 18-04 Guidance

Term Definition Source Safety-Related SSCs (SR SSCs that are credited in the fulfilment of Required Safety Functions and are capable NEI 18-04 and NEI 21-07 SSCs) to perform their Required Safety Functions in response to any Design Basis Hazard Level.

Safety-Significant SSC An SSC that performs a function whose performance is necessary to achieve NEI 18-04 and NEI 21-07, adequate Defense-in-Depth or is classified as risk-significant (see Risk-Significant embellished for this SSC). The population of Safety-Significant SSCs is made up of SR SSCs and NSRST document SSCs.

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Change Control Scope and Process for a Reactor Licensed Appendix B in Accordance with the NEI 18-04 Guidance

Table B-2. Corresponding Terms NEI 96-07 Change Control for an LMP -Based Affirmative Safety Case Accident Previously Evaluated in the FSAR (As Updated) The concept is the same for an LMP-based affirmative safety case.

Accident previously evaluated in the FSAR (as updated) means a However, Chapters 6 and 15 have a different meaning for an design basis accident or event described in the UFSAR including advanced reactor following the SAR guidance in NEI 21-07. The term accidents, such as those typically analyzed in Chapters 6 and 15 of the typically analyzed in Chapters 6 and 15 of the SAR corresponds to UFSAR, and transients and events the facility is required to withstand all LBEs (AOOs, DBEs, BDBEs, and DBAs) for a reactor that uses the such as floods, fires, earthquakes, other external hazards, anticipated NEI 18-04 methodology.

transients without scram (ATWS) and station blackout (SBO).

In LMP, event-sequences are evaluated in contrast to accidents in a The term accidents refers to the anticipated (or abnormal) traditional SRP-based approach.

operational transients and postulated design basis accidents that are analyzed to demonstrate that the facility can be operated without undue risk to the health and safety of the public.

Design Function For the purpose of addressing change control in a reactor with an Design functions are UFSAR-described design bases functions and LMP-based affirmative safety case, design functions are considered to other SSC functions described in the UFSAR that support or impact be Required Safety Functions per NEI 21-07 SAR Section 5.2, risk-design bases functions. Implicitly included within the meaning of significant functions per NEI 21-07 SAR Section 5.5.1, or a safety design function are the conditions under which intended functions function required for adequate DID in NEI 21-07 SAR Section 5.5.2.

are required to be performed, such as equipment response times, process conditions, equipment qualification, and single failure.

Design bases functions are functions performed by systems, structures, and components (SSCs) that are (1) required by, or otherwise necessary to comply with, regulations, license conditions, orders, or technical specifications, or (2) credited in licensee safety analyses to meet NRC requirements.

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Change Control Scope and Process for a Reactor Licensed Appendix B in Accordance with the NEI 18-04 Guidance

NEI 96-07 Change Control for an LMP -Based Affirmative Safety Case SSCs Important to Safety For the purpose of addressing change control in a reactor with an The term malfunction of an SSC important to safety refers to the LMP-based affirmative safety case, safety-related SSCs and NSRST failure of structures, systems and components (SSCs) to perform their SSCs, taken together, are considered to be equivalent to SSCs intended design functionsincluding both non-safety-related and important to safety.

safety-related SSCs.

Thus, an important safety SSC is one that carries out a design function (see above), but is not necessarily safety-related.

Methods of Evaluation The concept is the same for an LMP-based affirmative safety case.

Methods of evaluation means the calculational framework used for However, in the NEI 18-04 methodology, many of the LBEs are evaluating behavior or response of the facility or an SSC. evaluated as part of the plant PRA. For such methods of evaluation, changes are controlled by the Non-LWR PRA Standard, and as such, those methods of evaluation are outside the scope of 10 CFR 50.59.

Updated Final Safety Analyses Report (UFSAR) The concept is the same for an LMP-based affirmative safety case.

UFSAR refers to the safety analysis report (SAR) of a plant that has (i) However, the more general term SAR is often used in the NEI 18 -04 received its operating license and (ii) updated to reflect the current and NEI 21-07 guidance documents, which were written with state of knowledge. applicants in mind rather than licensees. Where the term SAR is used in this document, it can be interpreted as UFSAR.

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