ML25168A015
| ML25168A015 | |
| Person / Time | |
|---|---|
| Site: | Electric Power Research Institute |
| Issue date: | 07/15/2025 |
| From: | Licensing Processes Branch |
| To: | |
| References | |
| EPRI TR 3002025288, EPID L-2023-TOP-0031 pre-app, EPID L-2023-TOP-0045 pre-fee, EPID L-2023-NTR-0008 post-fee | |
| Download: ML25168A015 (39) | |
Text
Enclosure 7/15/2025 Safety Evaluation for a Topical Report Summary Information Topical Report No.:
Electric Power Research Institute Technical Report No. 3002025288 Topical Report
Title:
Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components Sponsor:
Electric Power Research Institute (EPRI)
Summary of Request:
EPRI requested the U.S. Nuclear Regulatory Commission to review and approve EPRI Technical Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components, June 2023 (EPRI 2023a) Applicability: All plants licensed under the Atomic Energy Act of 1954, as amended, implementing Title 10 of the Code of Federal Regulation (10 CFR)
Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, programs Submittal:
August 17, 2023, ADAMS Accession No. ML23234A266 (EPRI 2023a)
Supplements:
November 30, 2023, ADAMS Accession No. ML23334A210 (EPRI 2023b)
June 28, 2024, ADAMS Accession No. ML24180A016 (EPRI 2024a)
April 30, 2025, ADAMS Accession No. ML25121A201 (EPRI 2025a)
EPID No.:
L-2023-NTR-0008 Principal Contributors to Safety Evaluation Mihaela Biro, Senior Reliability and Risk Analyst, PRA Licensing Branch A (APLA),
Division of Risk Assessment (DRA), Office of Nuclear Reactor Regulation (NRR)
Jeff Circle, Senior Reliability and Risk Analyst, APLA, DRA, NRR Stephen Cumblidge, Materials Engineer, Division of New and Renewed, Division of New and Renewed Licenses (DNRL), Piping and Head Penetrations Branch NRR David Gennardo, Reliability and Risk Analyst, APLA, DRA, NRR Dan Widrevitz, Materials Engineer, Vessels and Internals Branch, DNRL, NRR
ii Table of Contents Summary Information................................................................................................................... i Table of Contents........................................................................................................................ ii
- 1.
Introduction..................................................................................................................... 1 1.1 Description of Request................................................................................................. 1 1.2 Background.................................................................................................................. 1
- 2.
Evaluation Criteria........................................................................................................... 1 2.1 Applicable Regulations................................................................................................ 1 2.2 Mandated Licensing Basis Document Information....................................................... 2 2.3 NRC-Approved Precedents.......................................................................................... 2 2.4 Applicable Guidelines.................................................................................................. 3 2.5 Overview of 10 CFR 50.69 Categorization Process..................................................... 4 2.6 ANO-2, ASME Code Case N-660, and ASME Code Case N-752 Passive Categorization Processes....................................................................................................... 5 2.7 Description of Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components in EPRI TR 3002025288.................................................................. 6
- 3.
Technical Evaluation....................................................................................................... 7 3.1 Scope of Review.......................................................................................................... 7 3.2 Method of Review........................................................................................................ 8 3.2.1.
Traditional Engineering Evaluation...................................................................... 8 3.2.2.
Risk-informed Evaluation...................................................................................25
- 4.
Limitations and Conditions..............................................................................................32
- 5.
Conclusion.....................................................................................................................32
- 6.
References.....................................................................................................................32
- 7.
Abbreviations..................................................................................................................36
- 1.
INTRODUCTION 1.1 Description of Request By letter dated August 17, 2023 (EPRI 2023a), as supplemented on November 30, 2023 (EPRI 2023b), June 28, 2024 (EPRI 2024a), and April 30, 2025 (EPRI 2025a), Electric Power Research Institute (EPRI) submitted EPRI Technical Report (TR) 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components, to the U.S. Nuclear Regulatory Commission (NRC) for review and approval. The April 30, 2025 submittal provided an amended markup of the topical report. EPRI TR 3002025288 presents a methodology for categorizing pressure boundary components in support of Title 10 of the Code of Federal Regulation (CFR) Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, (10 CFR 50.69) applications.
By letter dated July 11, 2024 (NRC 2024a), the NRC staff accepted EPRI TR 3002025288 for review.
1.2 Background
The NRC staff issued one round of requests for additional information (RAIs) containing nine questions:
January 13, 2025, ADAMS Accession No. ML24352A469 (NRC 2025a)
The NRC staff performed an audit to support its review:
NRC audit plan dated August 30, 2024, ADAMS Accession No. ML24241A160 (NRC 2024b)
NRC audit report dated June 12, 2025, ADAMS Accession No. ML25147A118 (NRC 2025b)
- 2.
EVALUATION CRITERIA 2.1 Applicable Regulations 10 CFR Part 50, Domestic licensing of production and utilization facilities (10 CFR Part 50) 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors (10 CFR 50.69)
The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems, and components (SSCs) subject to special treatment requirements. Special treatment refers to those requirements that provide increased assurance (beyond normal industry practices) that SSCs will perform their design-basis functions. For SSCs with a function that is categorized as low safety significant (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs with a function determined to be high safety significant (HSS), requirements may not be changed.
Section 50.69 of 10 CFR contains requirements regarding how a licensee categorizes SSCs using a risk-informed process; adjusts treatment requirements consistent with the relative
2 significance of the SSC; and manages the process over the lifetime of the plant. A risk-informed process is employed to determine the safety significance of SSCs and assign each into one of four Risk-Informed Safety Class (RISC) categories.
SSC categorization does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility. Instead, 10 CFR 50.69 enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained and may be enhanced. Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative, risk-informed approach to treatment that provides a reasonable level of confidence that these SSCs will satisfy functional requirements.
Implementation of 10 CFR 50.69 allows licensees to improve focus on HSS equipment.
2.2 Mandated Licensing Basis Document Information Paragraph 50.69(f)(2) of 10 CFR requires licensees to update their final safety analysis report to reflect which systems have been categorized.
2.3 NRC-Approved Precedents EPRI TR-112657, Revised Risk-Informed Inservice Inspection Evaluation Procedure,"
Revision B-A, February 10, 2000 (EPRI 2000a)
American Society of Mechanical Engineers (ASME) Code Case N-660, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement ActivitiesSection XI, Division 1" (ASME 2005a), approved in NRC Regulatory Guide (RG) 1.147, Revision 21, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, March 2024 (NRC 2024c)
NRC Approval of Arkansas Nuclear One (ANO) Unit 2, Request for Alternative from Certain Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (NRC 2021a)
NRC Approval of ANO, Unit 2, Request for Alternative ANO2-R&R-004, Revision 1, Request to use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (NRC 2009a)
NRC Issuance of Donald C. Cook Nuclear Plant, Units 1 and 2, Amendments Regarding Adoption of 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (NRC 2025c)
NRC Authorization of Grand Gulf Nuclear Station, Unit 1; River Bend Station, Unit 1; and Waterford Steam Electric Station, Unit 3, of Alternative to Use EN-RR-01 Concerning Proposed Alternative to Adopt Code Case N-752 (NRC 2024d)
NRC Authorization of NextEra Fleet - Proposed Alternative FRR 23-01 to use ASME Code Case N-752-1, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems Section X1, Division 1 (NRC 2024e)
3 NRC Authorization of Oconee Nuclear Station, Units 1, 2, and 3, Alternative to Use RR-22-0174, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1 (NRC 2023a)
NRC Authorization of Joseph M. Farley Nuclear Plant, Units 1 and 2; Edwin I. Hatch Nuclear Plant, Units 1 and 2; and Vogtle Electric Generating Plant, Units 1 and 2 - Proposed Alternative to use ASME Code Case N-752 (NRC 2025d)
NRC Approval of Seabrook Station Request for Alternative to Use ASME Code Case N-716 to Implement Risk-Inservice Inspection (RI-ISI) Program (NRC 2012a)
NRC Issuance of Vogtle Electric Generating Plant, Units 1 and 2, Amendments Regarding the Use of 10 CFR 50.69 (NRC 2014a)
ASME Code Case N-716-3, Alternative Classification and Examination RequirementsSection XI, Division 1 (ASME 716-3), approved in NRC RG 1.147, Revision 21, (NRC 2024c) 2.4 Applicable Guidelines RG 1.147, Revision 21, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, March 2024 (NRC 2024c)
RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, May 2018 (NRC 2018)
RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, December 2020 (NRC 2020)
RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, May 2006 (NRC 2006)
Nuclear Energy Institute (NEI) NEI 00-04 Technical Report, Revision 0, "10 CFR 50.69 SSC Categorization Guideline," July 31, 2005 (NEI 2005)
NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs
[probabilistic risk assessments] in Risk-Informed Decisionmaking (NRC 2017a)
NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, December 2010 (NRC 2010a)
NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, Revision 0 (NRC 2017b)
ASME / American Nuclear Society (ANS), Addenda to ASME/ANS RA-S-2008, Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009 (ASME 2009a)
Nuclear Management and Resources Council, Inc., Guidelines for Industry Actions to Assess Shutdown Management, NUMARC 91-06, December 1991 (NUMARC 1991)
4 2.5 Overview of 10 CFR 50.69 Categorization Process Paragraph 50.69(c) of 10 CFR requires licensees to use an integrated decisionmaking process to categorize safety-related and non-safety-related SSCs according to the safety significance of the functions they perform. They are placed into one of the following four RISC categories:
RISC-1:
Safety-related SSCs that perform safety significant functions RISC-2:
Non-safety-related SSCs that perform safety significant functions RISC-3:
Safety-related SSCs that perform LSS functions RISC-4:
Non-safety-related SSCs that perform LSS functions The SSCs have functions that are HSS or LSS, and they are classified accordingly. For SSCs with HSS functions (i.e., RISC-1 and RISC-2 SSCs), 10 CFR 50.69 maintains current regulatory requirements for special treatment, that is, all existing special treatment requirements remain in force. In addition, 10 CFR 50.69(d)(1) requires that the licensee or applicant shall ensure that RISC-1 and RISC-2 SSCs perform their functions consistent with the categorization process assumptions by evaluating treatment being applied to these SSCs to ensure that it supports the key assumptions in the categorization process that relate to their assumed performance. For SSCs with LSS functions, licensees may implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2). For RISC-3 SSCs, licensees may replace special treatment requirements with an alternative treatment approach that meets 10 CFR 50.69(d)(2). For RISC-4 SSCs, 10 CFR 50.69 does not impose new treatment requirements.
Section 50.69(b)(3) of 10 CFR states that the Commission will approve a licensees implementation of this section by issuance of a license amendment if the Commission determines that the categorization process satisfies the requirements of 10 CFR 50.69(c).
As stated in 10 CFR 50.69(b), after the NRC approves an application for a license amendment, a licensee may voluntarily comply with 10 CFR 50.69 as an alternative to compliance with the following requirements for LSS SSCs:
The specified portion of 10 CFR 50.46a(b)
(iii) 10 CFR 50.49 (iv) 10 CFR 50.55(e)
(v)
Specified requirements of 10 CFR 50.55a (vi) 10 CFR 50.65, except for paragraph (a)(4)
(vii) 10 CFR 50.72 (viii) 10 CFR 50.73 (ix)
Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50 (x)
Specified requirements for containment leakage testing (xi)
Specified requirements of Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, to 10 CFR Part 100, Reactor site criteria The NRC staff typically reviews a licensees proposed SSC categorization process against the categorization process described in NEI 00-04 and reviews the acceptability of the licensees PRA for use in the application of the 10 CFR 50.69 categorization process against the guidance
5 in RG 1.200. The NRC staffs review uses the framework provided in RG 1.174, Revision 3, and RG 1.201, Revision 1.
Section 2.0 of NEI 0004 states that the categorization process includes the following eight primary steps:
(1) Assembly of Plant-Specific Inputs (2) System Engineering Assessment (3) Component Safety Significance Assessment (4) Defense-in-Depth Assessment (5) Preliminary Engineering Categorization of Functions (6) Risk Sensitivity Study (7) Integrated Decision Making Panel Review and Approval (8) SSC Categorization The passive categorization process falls under Step 3, Component Safety Significant Assessment, and the NRC staffs review of the passive categorization process would be documented in NRC staffs plant-specific 10 CFR 50.69 safety evaluations (SEs).
2.6 ANO-2, ASME Code Case N-660, and ASME Code Case N-752 Passive Categorization Processes Sections 4.0 and 5.1 of NEI 00-04 (NEI 2005) reference ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities, as an approach for addressing the pressure-retaining function (also referred to as passive or pressure boundary components) or the passive function of active components. The guidance in RG 1.201, Revision 1, clarifies that the version of ASME Code Case N-660 that is acceptable to the NRC staff for use in this application is the version identified in RG 1.147 (NRC 2024c),
subject to any conditions or limitations specified therein. Trial applications of this code case demonstrated it was conservative and essentially categorizes almost all passive equipment as HSS (NRC 2009a).
The ANO-2 risk-informed repair/replacement application (RI-RRA) was developed as a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure-retaining items and their associated supports (exclusive of ASME Code Class CC and MC items), by modifying the ASME Code Case N-660 methodology. The methods are based, in part, on the consequence of failure evaluation for RI-ISI applications contained in EPRI TR-112657, Revision B-A (EPRI 2000a). The ANO-2 RI-RRA methodology was initially approved for use in request for alternative, ANO2-R&R-004, Revision 1 (NRC 2009a). This methodology was later approved as the categorization method for passive components for the Vogtle 10 CFR 50.69 pilot application in the NRC SE dated December 17, 2014 (NRC 2014a). To date, all licensees who have submitted license amendment requests (LARs) to adopt 10 CFR 50.69 rely on the ANO-2 RI-RRA methodology.
Code Case N-752 was later developed based on the ANO-2 RI-RRA method for use in relief requests (i.e., as part of a proposed alternative to requirements in 10 CFR 50.55a). The categorization process in Code Case N-752 is the same as in the ANO-2 RI-RRA method, although the scope of its application is different in that it allows for individual items within a system to be categorized, whereas 10 CFR 50.69 requires categorization be performed for
6 entire systems and structures. Code Case N-752 has been approved in multiple relief requests, including those for ANO-1 and ANO-2 in 2021 (NRC 2021a) and Oconee in 2023 (NRC 2023a).
The ANO-2 RI-RRA and Code Case N-752 methods categorize passive pressure-retaining components, and their associated supports, based only on the consequence of failure (i.e., component failure is assumed with a probability of 1.0) and do not consider their individual initiating event (or failure rate) frequencies. In other words, the methods rely on the conditional core damage probabilities (CCDPs) and conditional large early release probabilities (CLERPs) associated with passive component ruptures. The methods also apply deterministic considerations consistent with riskinformed decisionmaking principles (e.g., defense-in-depth (DID)) in determining the final safety significance of the component.
Safety significance is generally measured by the initiating event frequency and the conditional probability which represents the consequence of a particular event, in this case, a passive boundary rupture. However, treatment requirements (including repair and replacement) affect the frequency of passive component failure. For materials issues, CCDP is commonly used as a metric to determine the appropriateness for reductions in safety margin. CCDP is used in EPRI TR-112657 and ASME Code Case N-660 to determine the safety significance of SSCs, as CCDP is not sensitive to the changes in the initiating event frequency that can be expected with reductions in safety margin. Categorizing the significance of rupture based on the consequence of failure alone is conservative compared to the more general measure (which includes rupture frequency). The categorization will not be affected by changes in frequency arising from changes to the treatment.
2.7 Description of Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components in EPRI TR 3002025288 The EPRI TR 3002025288 categorization methodology for pressure boundary components contains a set of prerequisites and a predetermined set of HSS systems/subsystems coupled with a plantspecific review for potentially risk-significant or high consequence outliers. The process consists of five phases as part of its implementation in a 10 CFR 50.69 program:
Phase 1: Prerequisites Phase 2: Predetermined HSS Passive SSCs Phase 3: Design and plant-specific search for preliminary HSS passive SSCs Phase 4: Sensitivity study and integrated decisionmaking panel (IDP) review and concurrence Phase 5: Performance monitoring Before implementing the categorization process described in TR 3002025288, Section 4.2, a licensee will need to ensure that the following prerequisites have been met:
Prerequisite 1: PRA acceptability of the internal events PRA model, including internal flooding Prerequisite 2: Integrity management for passive components, including:
7 A robust program that addresses localized corrosion A robust program that addresses flow-accelerated corrosion (FAC)
A robust program that addresses erosion Prerequisite 3: Evaluation of protective measures for Internal Flooding events Prerequisite 4: PRA configuration control program to ensure the PRA is updated and that it reflects the asbuilt/asoperated plant The methodology in TR 3002025288 uses a set of twelve criteria to categorize some passive SSCs and their associated supports as preliminarily HSS, with the remaining passive SSCs not categorized as preliminarily HSS by these criteria being categorized as preliminarily LSS. The first ten criteria consist of generic configurations and lists of SSCs that would be HSS, depending on the plant type. The eleventh criterion is the only part of the methodology that is plant-specific and based on the quantitative results (risk metrics) of the internal events and internal flooding PRA models. The last criterion is specific to piping and component supports.
The categorization process used in TR 3002025288 could be considered an expansion of the process used in ASME Code Case N-716-3. EPRI TR 3002025288 relies, in part, on five of the criteria for HSS categorization in Section 2 of the Code Case. ASME Code Case N-716 was developed as a streamlined method of categorizing and prioritizing piping segments for RI-ISI programs. The goal of ASME Code Case N-716 was to simplify the process described in EPRI TR-112657, Revision B-A (EPRI 2000a). The new method in TR 3002025288 could also be considered a simplification and refinement or enhancement of the guidance in ANO-2 RI-RRA, as it uses less conservative risk criteria and removes or alters some of the deterministic criteria in that method.
- 3.
TECHNICAL EVALUATION 3.1 Scope of Review The NRC review and approval of EPRI TR 3002025288, as supplemented, is limited to the passive categorization methodology contained in EPRI TR 3002025288, Chapter 4. Any clarifications or limitations based on text in other chapters or RAI responses needed to support a regulatory finding are specifically identified in this SE. Chapters 1 through 3 mostly describe background and history and do not directly relate to the proposed method of categorizing SSCs, other than the discussion of applicability in Chapter 1. Chapter 5 is a review of systems in boiling water reactors (BWRs) and pressurized water reactors (PWRs) based on the criteria described in Chapter 4 and provides additional guidance which may be useful during implementation of EPRI TR 3002025288. Chapter 6 contains EPRIs assessment of the proposed methodology against the five key principles of risk-informed decisionmaking from RG 1.174, Revision 3. Chapter 7 is a summary and Chapter 8 includes the document references. In addition, the applicant provided substantial commentary and background in the supplements and RAI response. This evaluation highlights the portions of the RAI responses that underpin the NRC conclusions herein. Other portions of the RAI responses and supplements which do not affect the text in Chapter 4 should not be considered approved or unapproved based on this review.
Any guidance/discussion in EPRI TR 3002025288 not related to the passive categorization methodology and its incorporation as part of a 10 CFR 50.69 categorization program are not
8 within the scope of this review. For example, changes related to ASME Code classification of passive components (from ASME Code Class 1 to Class 2, etc.) are outside the scope of this review. Similarly, the NRC did not review or approve of any statements regarding the advantages of a particular methodology or benefits of one method versus another.
3.2 Method of Review The categorization method in EPRI TR 3002025288 was compared to the currently approved 10 CFR 50.69 methods and independently assessed to determine if it is technically adequate to provide a sufficient categorization process for passive SSCs. To be clear, the NRC staff evaluated the acceptability of this alternative method based on compliance with the requirements in 10 CFR 50.69. However, for additional guidance in evaluating specific elements of the method, the staff also compared the method to the risk categorization guidelines in NEI 00-04, as endorsed in RG 1.201, Revision 1, the expected results of its application with precedents for the application of the ANO-2 RI-RRA method, and with the generic risk-informed decisionmaking guidelines established in RG 1.174. This approach is similar to how the ANO-2 RI-RRA method was assessed against ASME Code Case N-660 (NRC 2009a).
The review of the methodology in Chapter 4 of EPRI TR 3002025288 against the ANO-2 RI-RRA method could be described as a traditional engineering evaluation and is presented in Section 3.2.1 of this SE.
An acceptable approach for making risk-informed decisions about proposed changes to the licensing basis is to show that the proposed changes meet the five key principles of risk-informed decisionmaking stated in Section C of RG 1.174, Revision 3 (NRC 2018). These key principles are:
Principle 1:
The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption.
Principle 2:
The proposed licensing basis change is consistent with the DID philosophy.
Principle 3:
The proposed licensing basis change maintains sufficient safety margins.
Principle 4:
When proposed licensing basis changes result in an increase in risk, the increases should be small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants.
Principle 5:
The impact of the proposed licensing basis change should be monitored using performance measurement strategies.
The assessment of the EPRI TR method against the five key principles of risk-informed decisionmaking is presented in Sections 3.2.2.1 through 3.2.2.5 of this SE.
3.2.1.
Traditional Engineering Evaluation The traditional engineering evaluation below evaluates differences from the proposed methodology as compared to the use of the ANO-2 RI-RRA method as part of 10 CFR 50.69
9 categorization. The NRC staff compared EPRI TR 3002025288 with the methodologies outlined in EPRI TR-112657, Revision B-A (EPRI 2000a), NEI 00-04 (NEI 2005), RG 1.201 (NRC 2006),
ASME Code Case N-660 (ASME 2005a), and ANO2-R&R-004 (NRC 2009a). The NRC staff took additional care in reviewing where treatment of active and passive components diverge from precedent.
3.2.1.1 Applicability of the Methodology As discussed in the response to RAIs 5a and 5b (EPRI 2025a), the methodology in EPRI TR 3002025288 is limited to plants currently or previously licensed via the 10 CFR Part 50 framework and plants with renewed licenses under the 10 CFR Part 54 framework. The response expands the last paragraph in Chapter 1 (and Chapter 7) of TR 3002025288 to state:
To that end, this report provides an enhanced approach for categorizing pressure boundary components for use in 10 CFR 50.69 applications. This methodology is based on decades of experience with risk-informing the pressure boundary, currently focused on plants licensed under 10 CFR [Part] 50 and plants with renewed licenses under 10 CFR [Part] 54. The methodology in this report is not currently applicable beyond the scope of plants for which the existing experience was used as part of the basis for methodology development.
The NRC staff notes that licensees of, and applicants for, dissimilar designs would need to provide design-specific justification as part of an application to use the methodology in TR 3002025288. For example, applicants or licensees seeking to reference this approval for plants with operating or combined licenses after December 31, 2018, would need to demonstrate that their design is sufficiently similar to the designs whose operating experience underpins this methodology and that the associated industry experience derived failure rates/pipe rupture frequencies are applicable. Alternatively, applicants may demonstrate that Criteria 1 through 10 of this methodology encompass a similar degree of the DID and risk aspects for their design as this methodology does for the designs considered in generating this methodology. The NRC staff notes that the risk metrics in Criterion 11 would also need to be justified for the newer design or alternate thresholds may need to be proposed. Overreliance on the application of Criterion 11 is insufficient to demonstrate the applicability of this methodology to a newer design.
As discussed in the response to RAI 2b (EPRI 2025a), passive categorization will continue to follow the applicable guidance outlined in NEI 00-04, with the addition of the prerequisites, the predetermined set of HSS components (Criteria 1 through 10) and the plant-specific review for risk significant components (Criterion 11) in Chapter 4 of EPRI TR 3002025288. The response goes on to provide text for EPRI TR 3002025288, Section 4.5, NEI 00-04 Integrated Decision Making Panel Guidance, which explains that, as required by Section 9.2 of NEI 00-04, the IDP is responsible for reviewing candidate HSS and LSS assignments and determining the final HSS and LSS assignment; and then identifies other guidance for the IDP in NEI 00-04.
Section 4.5 then provides additional requirements for the IDP when using the methodology in EPRI TR 3002025288.
Based on the above, and the discussion in EPRI TR 3002025288 which contains numerous references to 10 CFR 50.69 and NEI 00-04, it is understood that the use of this method is
10 contingent upon the use of NEI 00-04 for a 10 CFR 50.69 categorization process and is only applicable to existing plants which were used as the basis for the methodology.
3.2.1.2 Prerequisites (TR Section 4.1)
Chapter 4.1 of EPRI TR 3002025288 describes a set of prerequisites before a licensee is allowed to use the proposed categorization method.
Sections 4.1.2 and 4.1.3 of EPRI TR 3002025288 are a new approach and have no equivalent in the basis for ANO-2 RI-RRA.
TR Subsection 4.1.1, Prerequisite 1: PRA Technical Adequacy (Pressure Boundary Failures)
The methodology in EPRI TR 3002025288 requires that the plant must have a robust internal events PRA, including internal flooding, that addresses failure of all pressure boundary components (such as main steam line breaks, main feedwater line breaks, internal flooding events, and interfacing system loss-of-coolant accidents (LOCAs)). The plant-specific PRA must be sufficient to support the LAR approval process, including consideration of PRA assumptions and sources of uncertainty. Paragraph 50.69(c)(1)(i) of 10 CFR requires, in part, that the PRA be of sufficient quality and level of detail to support the categorization process, and it must be subjected to a peer review process assessed against a standard or set of acceptance criteria endorsed by the NRC. Paragraph 50.69(b)(2)(iii) of 10 CFR requires that the results of the PRA review process conducted to meet 10 CFR 50.69(c)(1)(i) be submitted as part of the application.
Prerequisite 1 in EPRI TR 3002025288 states that this can include full-scope peer review of the internal events and internal flooding PRA against RG 1.200, Revision 2 (NRC 2009b), as well as a gap assessment of earlier peer reviews of the internal events and internal flooding PRA against RG 1.200, Revision 2.
The NRC staff notes that RG 1.200 has been updated to Revision 3 (NRC 2020). However, the use of RG 1.200, Revision 2 (NRC 2009b), to demonstrate PRA acceptability (also known as PRA technical adequacy) remains acceptable, as discussed in Revision 3 of the RG.
In the response to RAI 3a, EPRI expanded on Prerequisite 1 to ensure the importance of properly considering and addressing uncertainty is underscored by adding the following considerations:
The analyst must review key assumptions and sources of model uncertainty in the context of this application. For example, prior to using the enhanced categorization methodology, any non-conservatisms or the use of methods not commonly accepted for risk-informed applications must be reviewed to determine their impact, if any, on the risk-informed categorization of the pressure boundary.
The NRC staff finds that the above considerations in Prerequisite 1 are consistent with the requirements for submitting an application to adopt 10 CFR 50.69 and practices for assessing the impact of uncertainty on an application. Additional discussion on PRA acceptability and uncertainty is provided in Section 3.2.2.4 of this SE.
11 TR Subsection 4.1.2, Prerequisite 2: Integrity Management EPRI TR 3002025288, Section 4.1.2, describes programs related to pressure boundary integrity as necessary prerequisites to the methodology supporting compliance with 10 CFR 50.69(d)(2).
Specifically, 10 CFR 50.69(d)(2) requires that applicants shall ensure, with reasonable confidence, that RISC-3 SSCs remain capable of performing their safety-related functions under design-basis conditions through inspection and testing; and corrective actions.
The methodology includes systems/subsystems managed within RI-ISI programs, such as via ASME Code Case N-716-3. Components assigned a final HSS determination by the IDP would remain within the scope of a licensees RI-ISI program. The NRC staff notes that this means that treatment of these HSS components would remain consistent with the basis for approval of said programs and not be superseded by EPRI TR 3002025288 methodology.
The methodology lists several programs as a robust mechanism for addressing the 10 CFR 50.69(d)(2) requirements for systems typically outside the scope of an ISI program (RI-ISI or otherwise), specifically addressing:
Localized corrosion FAC Erosion These are degradation mechanisms that primarily (but not exclusively) effect piping and slowly degrade material until repair/replacement is warranted to maintain safety margins. Managing these forms of degradation enables the reliability and availability of associated SSCs.
The robustness of the listed programs is described in the methodology through providing examples of a variety of EPRI and NEI reports such as EPRI TR-103403, Service Water System Corrosion and Deposition Sourcebook (EPRI 1994), etc. In clarifying the intent of this methodology section during the audit (NRC 2025b), and in response to RAI 6, the applicant presented extended examples of robust programs that committed to the use of specified treatments under similar reports.
The applicant clarified that the robustness of any program alternatives would be determined in relationship to the example programs (e.g., demonstration of equivalent robustness through comparison to the example programs). These examples describe the degree of program robustness consistent with the methodology, as explained through a clarifying revision (EPRI 2025b) to the methodology, In the LAR submittal for implementing the enhanced passive categorization, the licensee shall identify the programs for addressing localized corrosion, flow-accelerated corrosion, and erosion. The following bullets provide examples of industry guidance, if other guidance is used, the licensee shall provide justification to demonstrate robustness.
Further, the methodology was revised to clarify that later alterations to such programs would require consideration under 10 CFR 50.69(e) and may result in updated categorization and treatment of SSCs.
12 The NRC staff understand existing programs concerning the listed corrosion and erosion mechanisms are part of site management and maintenance programs ensuring the adequate operation of the site. In addition, these programs support proactive planning to ensure the reliability of and manage maintenance for SSCs subject to the degradation mechanisms listed in the methodology. Many of these programs are included in licensing applications such as license renewal applications, where these programs are often also placed under 10 CFR Part 50, Appendix B, requirements.
The NRC staff notes that use of EPRI TR 3002025288 does not supersede actions, activities, commitments, etc. made in other licensing actions (e.g., renewed licenses). Specifically, operating programs for which 10 CFR Part 50, Appendix B, has been relied upon should remain under Appendix B. Changes to such programs would need to be made using the appropriate change control processes (e.g., 10 CFR 50.59 for final safety analysis report updates and 10 CFR 50.54 for quality assurance program changes).
The use of permissive terms in the methodology regarding these prerequisite programs (e.g., should) does not mean that a lack of such programs would be acceptable under the methodology. This was clarified during the audit, namely that a lack of equivalently robust programs would leave the acceptability of the application open. The methodology relies on programs consistent with the examples as necessary prerequisites to the use of the methodology. Consequently, and as clarified in response to RAI 6, the NRC staff notes that deviation from the given examples would require justification by the applicant as described in the revised text quoted above. Without such justification the applicant would not be within the bounds of the methodology. Future alterations to said programs would require evaluation under 10 CFR 50.69(e), consequently ensuring alterations are appropriately evaluated relative to the methodology.
Paragraph 50.69(b)(2)(iv) of 10 CFR requires that a LAR to adopt 10 CFR 50.69 include a description of, and basis for acceptability of, the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1)(iv). These evaluations must include the potential impacts from known degradation mechanisms for both active and passive functions. The NRC staff reviewed the example programs, and their relationship to the rest of the methodology, to confirm that findings under 10 CFR 50.69(c)(1)(iv) would be supported by the example programs. The NRC staff confirmed that the combination of the robustness of management of localized corrosion, FAC, erosion through the example programs, and the normal ISI programs (including RI-ISI) would appropriately detect and manage service-induced degradation in combination with the rest of the methodology. This finding is further supported by the fact that many of these programs are 10 CFR Part 50, Appendix B, programs for plants with renewed licenses under 10 CFR Part 54; thus, enhancing the quality and reliability of application of said example programs.
TR Subsection 4.1.3, Prerequisite 3: Protective Measures for Internal Flooding Events Prerequisite 3 includes guidance on protective measures for internal flooding events (i.e., floor drains, flood alarm equipment, and barriers). The prerequisite states that such protective measures should not be categorized as LSS unless additional evaluations have been conducted to show that loss of these measures or a subset of them will not invalidate the HSS determination provided in EPRI TR 3002025288, Section 4.2. For example, if a submarine door has been credited in preventing a flood from exiting one flood zone and entering another, that submarine door will be considered HSS or left uncategorized unless an evaluation has shown
13 that loss of the door will not significantly increase plant risk (i.e., exceed the risk criteria in Criterion 11). It is noted that structural barriers are not considered to be pressure boundary components and are not categorized as part of this methodology but could be categorized as part of an overall 10 CFR 50.69 program.
The NRC staff finds this prerequisite acceptable as it addresses protective measures that may not be typically modeled in a licensees PRA and ensures that they are not assigned preliminary LSS without additional justification. Application of this prerequisite is consistent with the requirements of 10 CFR 50.69(c)(1)(ii) to determine SSC functional importance using an integrated, systematic process, including addressing SSCs not modeled in the PRA.
TR Subsection 4.1.4, Prerequisite 4: Reflect the As-Built/As-Operated Plant Section 4.1.4 was added to EPRI TR 3002025288, by letter dated November 30, 2023 (EPRI 2023b), to include additional Prerequisite 4 to ensure the PRA reflects the as-built/as-operated plant. This prerequisite requires that:
A PRA Configuration Control Program shall be in place in accordance with the ASME/ANS PRA Standard that maintains the PRA, and any supplementary analysis, to reflect the asbuilt/asoperated plant.
A feedback and adjustment process shall be in place to review changes to the plant, operational practices, applicable plant and industry operational experience, and as appropriate, update the PRA model and if necessary, the categorization and treatment processes.
This review shall meet the required update periodicity requirements for input information, such as the PRA update requirements for the underlying information used as input in the methodology implementation.
Paragraph 50.69(c)(1)(ii) of 10 CFR requires that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.
Paragraph 50.69(e) of 10 CFR requires that periodic updates to the licensees PRA and SSC categorization be performed. Changes over time to the PRA and to the SSC reliabilities are inevitable, and such changes are recognized by the 10 CFR 50.69(e) requirement for periodic updates. Maintaining change control and conducting periodic reviews provides confidence that all aspects of the program reasonably reflect both the current as-built, as-operated plant configuration and applicable plant and industry operational experience as required by 10 CFR 50.69(c)(1)(ii). The NRC staff finds that the requirements in Prerequisite 4, when applied in conjunction with the requirements NEI 00-04, as endorsed by RG 1.201, Revision 1, are acceptable and are consistent with the requirements in 10 CFR 50.69(c)(1)(ii) and (e).
Additional discussion on feedback and process adjustment is provided in Section 3.2.2.5 of this SE.
3.2.1.3 Scope of Passive Categorization Methodology In response to RAI 2a, EPRI revised TR 3002025288, Section 4.6, Alternative Treatment Requirements Under 10 CFR 50.69(d)(2). This section states, in part: This enhanced methodology for passive categorization requires an evaluation of the pressure retention function
14 of all systems. In this approach the pressure boundary, function is treated as a system for 10 CFR 50.69 categorization and alternative treatment purposes, whereas the traditional passive methodology is applied on a system-by-system basis.
Paragraph 50.69(c)(1)(v) of 10 CFR requires the SSC categorization process to [b]e performed for entire systems and structures, not for selected components within a system or structure. As described in the statements of consideration (69 FR 68008 (November 22, 2004)):
This required scope ensures that all safety functions associated with a system or structure are properly identified and evaluated when determining the safety significance of individual components within a system or structure and that the entire set of components that comprise a system or structure are considered and addressed.
The statements of consideration also note that this requirement should be understood to exclude entire support systems (e.g., if System A is categorized as RISC-3, but is dependent on System B components which in turn have been categorized as RISC-1, then System A is understood not to include the System B components and is not to be categorized as RISC-1).
The comment responses to Issue Nos. c-6, c-13, and c-29 in SECY-04-0109 - Response to Public Comments to Final Rule 10 CFR 50.69 (NRC 2004a), provide additional insights as to how system boundaries are defined and the purpose of the requirement in 10 CFR 50.69(c)(1)(v). The NRC response to Issue No. c-29 states, in part: system boundaries are to be defined by the licensee and should be consistent with the PRA used in the categorization process. The NRC response to Issue No. c-6, states, in part: The primary reason that 10 CFR 50.69 requires the categorization to be performed for entire systems and structures is to ensure that all the functions (which are primarily a system-level attribute) for a given SSC within a given system or structure are appropriately considered for each SSC in determining its safety significance. The system boundary definitions should be consistent with the PRA used in categorizing the SSCs and careful consideration should be given by the licensee to ensure all important functions are captured for SSCs, especially those that are common to multiple systems (e.g., tank discharge valve that feeds to multiple systems). The methodology for determining systems boundaries is left to the licensee recognizing these important constraints (i.e., drawing system boundaries in such a way as to break apart a system when viewed from a system functional standpoint would not meet this requirement).
For passive pressure-retaining components the only assigned function is to maintain the pressure boundary (e.g., a valve body is not assigned the active function associated with the valve). Treating the pressure boundary function as a system is consistent with 10 CFR 50.69 rule language in that there will be no other important functions that would escape categorization and appropriate assignment of safety significance. Licensees are allowed to define system boundaries and defining a system based on the pressure boundary function would also be analogous to defining a new support system.
The NRC staff notes that this approach could not be applied to redefining other active components based on a single function (e.g., a pumping system with all the pumps and no other equipment) because the active component functions are tied to the system functions and system boundaries should not be redefined in a manner that is not consistent with the PRA and apart a system when viewed from a system functional standpoint.
15 The passive categorization methodology in EPRI TR 3002025288 does not apply to active components. When categorizing a system that contains both active and passive (pressure boundary) components, active components (e.g., nonpressure-retaining functions) must follow the existing process for categorization in NEI 0004 which ensures that all safety functions are properly identified and categorized regarding their safety significance. The NRC staff understands that situations can exist where parts of a component can have active and passive functions (e.g., a valve internal can be categorized as HSS with the valve body being categorized as LSS (NRC 2014a)). In such a case, the NRC staff concludes that alternative repair/replacement activities can be applied to the LSS pressureretaining function of the component, and the active HSS function will continue to be maintained through appropriate practices.
Similarly, this approach allows active components to remain uncategorized while alternative treatments can be applied to LSS passive components. The NRC staff notes that this approach is consistent with approval of relief requests to use ANO-2 RI-RRA and ASME Code Case N-752 that allow passive categorization without categorizing the active function(s)
(NRC 2009a, NRC 2023a). This means that the alternative treatments for the LSS passive components cannot affect the active components/functions and the active component/function will continue to be maintained through appropriate practices.
Based on the above discussion, the NRC staff finds that the approach in EPRI TR 3002025288 to define the pressure boundary function as a new system for 10 CFR 50.69 categorization complies with the requirements in 10 CFR 50.69(c)(1)(v).
3.2.1.4 TR Section 4.2, Predetermined HSS Passive SSCs EPRI TR 3002025288 uses a set of twelve criteria to categorize passive components and their associated supports. A component satisfying any one of these criteria will be categorized as preliminary HSS, with all other SSCs not categorized as preliminary HSS by these criteria being categorized as preliminary LSS. Criterion 11 is the only part of the methodology that is plant-specific and based on the results (risk importance measures) of the internal events and internal flooding PRA model(s).
The categorization process proposed in EPRI TR 3002025288 to categorize SSCs is an expansion of the process used in ASME Code Case N-716. ASME Code Case N-716 was developed as a streamlined method of categorizing and prioritizing piping segments for RI-ISI programs. The goal of ASME Code Case N-716 was to simplify the complex process described in EPRI TR-112657, Revision B-A (EPRI 2000a).
EPRI TR 3002025288 expands the five criteria in ASME Code Case N-716 for use in categorizing SSCs for 10 CFR 50.69. Criteria 5 through 10 were added to EPRI TR 3002025288 as ASME Code Case N-716 only covers piping segments, with a focus on ASME Code Class 1 and 2 piping segments. Criteria 5 through 10 cover items such as vessels, heat exchangers, the ultimate heat sink, and other safety systems. Criterion 11 applies risk criteria and expands on Criteria 5 from ASME Code Case N-716.
As described above, EPRI TR 3002025288 proposes several concepts which could be considered novel for characterizing passive pressure-retaining SSCs for 10 CFR 50.69. Each of these concepts were evaluated to determine if they adequately assigned SSCs as HSS or LSS
16 and how the method described in EPRI TR 3002025288 compares to the ANO precedent.
These concepts include:
Creating a predetermined list of SSCs that would be designated as preliminary HSS based on their function, size, and the presence of redundant trains or systems. The NRC staff notes that there is precedent applying safe shutdown equipment lists as part of the qualitative categorization process for external hazards.
Considering the probability of SSC failure as part of the risk analysis by using core damage frequency (CDF) and large early release frequency (LERF) in addition to CCDP and CLERP in the risk criteria. The NRC staff notes that there is precedent based on criteria in Supporting Requirement IFQU-A3 of the endorsed ASME PRA Standard (ASME2009a) for considering both CDF and CCDP to screen flood areas from the internal flooding PRA.
Assuming any SSC not determined to be preliminary HSS by the criteria is preliminary LSS.
Predetermined Criteria Criterion 1 describes all ASME Code Class 1 SSCs as HSS. It then provides some guidance as to what Class 1 SSCs may be eligible to be reclassified as something other than Class 1 outside of the 10 CFR 50.69 process. This criterion is consistent with the ANO precedent in that all ASME Code Class 1 SSCs are HSS. Actions taken by a licensee outside of the 10 CFR 50.69 process are outside the scope of this review and would be governed by the appropriate change control processes.
Criteria 2 through 4 have been used for RI-ISI categorization for ASME Code Case N-716-3 (NRC 2024c). These criteria were developed by reviewing the results of the application of the EPRI RI-ISI evaluation method described in EPRI TR-112657, Revision B-A (EPRI 2000a), to several nuclear power plants. These criteria were developed to categorize the same piping welds as HSS as the piping welds commonly determined to be HSS using EPRI TR-112657 Revision B-A (EPRI 2000a).
Criterion 5 is used to designate SSCs in the ultimate heat sink (UHS) flow path that could result in a total loss of heat sink function as preliminary HSS. The HSS SSCs include any SSCs that would directly cause the loss of the UHS. The response to RAI 1a (EPRI 2025a) stated that Criterion 5 would be clarified to include any SSCs that, if they failed, would indirectly cause a loss of ultimate heat function through events such as internal flooding.
Criterion 6 designates tanks and vessels that support the emergency core cooling system, along with their respective piping, up to the first isolation valve as HSS. This designates SSCs that are important for DID evaluations of SSCs.
Criterion 7 covers the condensate storage tank for the emergency/auxiliary feedwater system.
These systems can be designated as LSS if there is an alternate and independent source of water for the system.
Criterion 8 covers SSCs for lowvolume, intermediatesafety systems, such as component cooling water that contain multiple trains. These SSCs are designated as HSS if the failure of
17 the SSC would result in the total loss of all trains. The response to RAI 1a (EPRI 2025a) stated that Criterion 8 would be clarified to include any SSCs that, if they failed, would indirectly cause a loss of both trains through events such as internal flooding.
Criterion 9 designates SSCs in heat exchangers as HSS if the failure of the SSC could allow reactor coolant to bypass primary containment. This criterion makes SSCs that can result in radioactive release HSS even if they are not modeled in the plant PRA. Criterion 9 was changed in the 2nd supplement submitted by EPRI (EPRI 2024a) to include additional guidance and references to the ANO-2 precedent (NRC 2009a) and EPRI guidelines for repair/replacement activities (EPRI 2011).
Criterion 10 directs licensees to evaluate other heat exchangers to determine which SSCs could affect multiple systems if they failed. If so, the methods described in References 5 and 6 of the TR are to be used to determine if the SSC is HSS or LSS.
Plant-Specific Criteria In the response to RAI 1a, EPRI revised Criterion 11 to categorize SSCs as preliminary HSS as follows:
Any piping or component, including piping segments or components grouped or subsumed within existing plant initiating event groups (main feedwater breaks inside containment; main steam line breaks outside containment; service water flooding events; interfacing system LOCAs; failures of non-Class 1 RCPB connections, such as instrumentation lines; and so on) whose contributions to:
CDF is greater than 10-06/year, or LERF is greater than 10-07/year, or scenarios where the:
CCDP is greater than 10-02, or CLERP is greater than 10-03.
Note: The 1x10-02 (CCDP)/1x10-03 (CLERP) values are similar to EPRI TR-112657, Revision B-A, and deterministic single failure criteria, seismic margin analysis, and fire protection (10 CFR Part 50, Appendix R) in that having a success path results in adequate protection for low frequency events.
This criterion is applied to a plant-specific PRA model that includes pressure boundary failures (for example, pipe whip, jet impingement, spray, and inventory losses).
For comparison, when applying the quantitative indices in Table I-5, the ANO-2 RI-RRA methodology identifies components with a CCDP/CLERP > 10-4/10-5 as high consequence which would then lead to a categorization of HSS.
The NRC staff notes that the risk measures of CCDP/CLERP are often used as surrogates to gauge the level of DID remaining after failure of a specified SSC. Additionally, as discussed in Section 2.6 of this SE, CCDP/CLERP are also relatively insensitive to changes in treatment,
18 which would only potentially impact the failure frequency of components. When considered in isolation, the change in risk metrics relative to the current CCDP/CLERP thresholds is therefore considered to be a reduction in conservatism as compared to ANO-2 RI-RRA methodology.
See Section 3.2.1.4, Additional discussion of DID and qualitative considerations, of this SE for additional discussion of the proposed CCDP/CLERP metrics and how they compare to precedents for more qualitative assessments of DID.
However, as discussed in the response to RAI 1a, the intent of the new methodology is to consider both CDF/LERF and CCDP/CLERP metrics to obtain more holistic insights on both frequency of occurrence and consequences of pipe ruptures in assessing risk, without unduly biasing the results towards overly conservative treatment of SSCs that are not significant to risk as measured by CDF and LERF alone.
The NRC staff acknowledges that the CDF and LERF metrics selected are consistent with guidance in ASME Code Case N-716-3 for identification of potentially risk significant piping segments when used in conjunction with the other four criteria in ASME Code Case N-716-3.
The NRC staff notes that risk acceptance guidelines in RG 1.174 are used to determine whether the increase in risk associated with a proposed change is small and consistent with the intent of the Commissions Safety Goal Policy Statement. Although a change in treatment requirements could impact the failure frequency of the associated passive component it would not necessarily result in a change in risk similar to the risk significance of the component itself.
As discussed in EPRI TR 3002025288 and the response to RAI 1a, the intent of Criterion 11 is to identify risk-significant outliers not assigned preliminary HSS by the first ten criteria. The inclusion of both CDF/LERF and CCDP/CLERP metrics in Criterion 11 ensures that plant-specific outliers are identified both in terms of risk and consequence. The NRC staff notes that this criterion would not be expected to capture and identify many if any additional SSCs that need to be designated as preliminary HSS, based on risk insights from previous reviews and review of the audit materials. Criterion 11 is useful as a backup to the first ten criteria but is not sufficient to identify HSS SSCs on its own. Therefore, the NRC staff finds the use of Criterion 11 acceptable for passive categorization only when it is used in conjunction with Criteria 1 through 10 per the process in Chapter 4 of EPRI TR 3002025288. See Section 3.2.2.4 of this SE for additional discussion of Criterion 11 and considerations related to uncertainty.
Clarification on the Application of Criterion 11 The NRC staff notes that Figure 3 of EPRI TR 3002025288 appears to indicate that if a criterion (i.e., 1 through 10) results in a component being assigned preliminarily HSS in Phase 2 that the next steps are unnecessary. However, as discussed in the responses to RAIs 7a and 7b, and explained in Section 4.2, the intent is that regardless of whether a pipe segment is determined to be HSS or LSS by Criteria 1 through 10, the pipe segment must still be assessed against Criterion 11 (to include CDF/LERF and CCDP/CLERP metrics). The NRC staff notes that if a passive component is determined to be HSS by Criteria 1 through 10, then there would be no change in categorization by assessing it against Criterion 11. Therefore, the NRCs approval of the methodology is not dependent on applying Criterion 11 to components that are already identified as preliminarily HSS.
19 Categorization of Supports Criterion 12 covers piping and component supports, including hangers and snubbers. This criterion is consistent with the ANO-2 RI-RRA precedent and its application to supports as discussed in the majority of LARs to adopt 10 CFR 50.69 in that it allows: (1) supports to remain uncategorized, and (2) if categorized, supports are assigned the same categorization (i.e., safety significance) as the highest ranked component they support. Criterion 12 adds an additional option whereby a combination of restraints or supports such that the LSS piping and associated SSCs attached to the HSS piping are included in the scope up to a boundary point that encompasses at least two supports in each of three orthogonal directions. This new option is consistent with guidance in the standard review plans for license renewal and subsequent license renewal and was previously approved on a plant-specific basis for D.C. Cook (NRC 2025c). Therefore, the NRC staff finds that the methodology for categorization of supports is acceptable and consistent with current guidance.
Discussion of Audit Findings During the audit (NRC 2025b), the fourteen original proposed criteria were compared to the risk information for the ASME Code Class 2 and 3 SSCs for a variety of PWR and BWR designs.
Based on the review, application of the ten predetermined criteria was found to be very consistent with the results of the ANO precedent for ASME Code Class 2 SSCs. This is in line with the use of ASME Code Case N-716 being similar to the results of EPRI TR-112657, Revision B-A (EPRI 2000a). While the audit found no SSCs that would be designated as HSS using Criterion 11, Table 7 provides some examples of when Criterion 11 would designate an SSC as preliminary HSS. A review of proposed alternatives to use ASME Code Case N-716 before it was incorporated by reference into RG 1.147 did find one case: Proposed Alternative for Palisades Nuclear Plant Relief Request Number RR 4-20 Proposed Alternative, Use of Alternate ASME Code Case N-716 (Entergy 2014a) where ASME Code Class 2 main steam piping welds had CDF contributions of greater than 10-6 per year and would be designated as HSS using Criterion 11.
The NRC staff finds that the group of SSCs classified as preliminary HSS by the ten criteria are substantially consistent with the comparable group categorized as HSS via the ANO precedent.
Correspondingly, the empirical derivation of the ten criteria from precedent results in a substantial portion of the categorizations supporting DID and risk considerations being appropriately categorized (e.g., as HSS). The process used to derive these ten criteria was only briefly described in the initial TR. Additional information was added in response to the RAIs, for example in the Premise column of TR Table 3 (e.g., consistent with LARs approved to date),
and in the response to RAI 5. The applicant clarified in RAI 5 that the basis of the methodology in EPRI TR 3002025288 is built on Generation II plant designs. The methodology for deriving the ten criteria is consequently inextricably bound to the designs and precedents used to derive the criteria. The NRC staff notes that this limits the extensibility of the methodology to new designs. Consequently, only such designs which are consistent with the precedents used to derive Criteria 1 through 10 may use this methodology without further basis. The audit did find a large number of ASME Code Class 3 SSCs that could have been categorized as LSS when applying the original ten predetermined criteria but were HSS using the ANO precedent. These SSCs are in high-risk flood zones with significant consequences for internal flooding with CCDPs greater than 1x10-4. In addition, these SSCs would also be designated as LSS using the initial quantitative risk Criteria 11 through 13 in the EPRI TR 3002025288. A review of the
20 documents provided during the audit indicated that application of the initial criteria could have resulted in SSCs being categorized as LSS under the circumstances described above if the internal flooding and external events were not factored in to the use of Criteria 5 and 8. The revisions to Criteria 5 and 8 in the response to RAI 1a (EPRI 2025a) were made to clarify that internal flooding events need to be taken into account.
EPRI TR 3002025288 gives implementation guidance for Criterion 1 and for Criteria 5 through 10. (Criteria 2 through 4 have been used extensively in RI-ISI applications for decades and thus require less discussion.) The implementation guidance describes the intent for some of the criteria and gives examples of SSCs that should be characterized as HSS. Additional technical information for each of the criteria is provided in Table 9 in Appendix A of EPRI TR 3002025288 that was originally given in the supplement dated June 28, 2024. This technical information can also be used to help licensees determine the appropriate SSCs to be HSS and facilitate consistent third-party review of the categorizations.
Based on the audit results and the revisions to EPRI TR 3002025288, the NRC staff finds that the twelve criteria and guidance given in EPRI TR 3002025288 provide a similar outcome as the ANO precedent for a light water reactor initially licensed to operate before December 31, 2018.
Use of EPRI TR 3002025288 for light water reactors licensed after December 31, 2018, would require additional technical information to show that the 12 criteria and guidance would provide an adequate categorization process.
Summary and Finding for HSS Passive SSCs Criteria 1 Through 10 Based on the above, the NRC staff finds that the application of Criteria 1 through 10 are consistent with the requirements of 10 CFR 50.69(c)(1)(ii) and (iii) and are suitable for use in applications, in concert with the other elements of the EPRI TR 300205288 methodology as discussed in this SE, to adopt or amend use of 10 CFR 50.69.
Additional Discussion of Defense-in-Depth (DID) and Qualitative Considerations The ANO-2 RI-RRA methodology contains additional requirements related to DID and qualitative considerations. For DID purposes, Section I-3.1.2(b) requires differences in the consequence rank between the use of Table I-2, Guidelines for Assigning Consequence Categories to Failures Resulting in System or Train Loss, which contains deterministic criteria based on the frequency of challenge, exposure time, and number of unaffected backup trains, and Table I-5, Quantitative Indices for Consequence Categories, which contains CCDP/CLERP limits for use with the PRA, to be reconciled. Similar tables are also used in ASME Code Cases N-660 and N-752. Additionally, Section I-3.1.2(b) requires postulated failures that lead to zero defense (i.e., no backup trains) be assigned a high consequence, which corresponds to HSS for 10 CFR 50.69 categorization.
The methodology in EPRI TR 3002025288 does not have an additional qualitative assessment like Table I-2 in the ANO-2 RI-RRA methodology. However, Criterion 11 in EPRI TR 3002025288 includes CCDP/CLERP limits (i.e., CCDP > 1x10-2 and ILERP > 1x10-3) which correspond to the assumed reliability for 1.0 trains in the ANO-2 RI-RRA method (i.e., one full train unavailability of approximately 1x10-2). This value is the same as the unavailability mean value in Table 3-10 of EPRI TR-112657, Revision B-A. Additionally, the use of CDF and LERF metrics in Criterion 11 accounts for the frequency of potential challenges. Furthermore, the
21 pre-determined set of HSS items identified via Criteria 1 through 10 of the methodology already account for the impact of reduced backup systems to provide key safety functions.
Finally, the NRC staff notes that original use of a qualitative assessment like Table I-2 predates the NRC guidance for PRA acceptability, which is contained in RG 1.200. The PRA models used in 10 CFR 50.69 applications are required to be of sufficient scope, level of detail, conformance with PRA technical elements, and plant representation, such that such a qualitative assessment of this nature would not be expected to provide any additional insights.
Therefore, the NRC staff finds that additional qualitative assessment to consider the frequency of challenge, exposure time, and number of unaffected backup trains, is not necessary, as the application of Criteria 1 through 11 in Chapter 4 of EPRI TR 3002025288 already accounts for these considerations in an acceptable manner.
Section I-3.2.2 of the ANO-2 RI-RRA methodology has additional categorization considerations.
Specifically, if any of the following ten conditions are answered FALSE, then HSS shall be assigned.
(1) Failure of the pressure-retaining function of the segment will not directly or indirectly (e.g., through spatial effects) fail a basic safety function.
(2) Failure of the pressure-retaining function of the segment will not prevent the plant from reaching or maintaining safe shutdown conditions; the pressure-retaining function is not significant to safety during mode changes or shutdown. Assume that the plant would be unable to reach or maintain safe shutdown conditions if a pressure-boundary failure results in the need for actions outside of plant procedures or available backup plant mitigative features.
(3) The pressure-retaining function of the segment is not called out or relied upon in the plant emergency/abnormal operating procedures or similar guidance as the sole means for successfully performing operator actions required to mitigate an accident or transient.
(4) The pressure-retaining function of the segment is not called out or relied upon in the plant emergency/abnormal operating procedures or similar guidance as the sole means for assuring long-term containment integrity, monitoring of post-accident conditions, or offsite emergency planning activities.
(5) Failure of the pressure-retaining function of the segment will not result in an unintentional release of radioactive material that would result in implementing offsite radiological protective actions.
The RISC process shall demonstrate that the DID philosophy is maintained. DID is maintained if:
(6) Reasonable balance is preserved among preventing core damage, preventing containment failure or bypass, and mitigating an offsite release.
(7) There is no overreliance on programmatic activities and operator actions to compensate for weaknesses in plant design.
22 (8) System redundancy, independence, and diversity are preserved commensurate with the expected frequency of challenges, consequences of failure of the system, and associated uncertainties in determining these parameters.
(9) Potential for common-cause failures is taken into account in the risk analysis categorization.
(10) Independence of fission-product barriers is not degraded.
In Attachment 2 of the second supplement (EPRI 2024a), and in response to RAI 3 (EPRI 2024b), EPRI explains how the methodology in Section 4 of EPRI TR 300205288 addresses these considerations.
For Questions 1 and 2, any pressure boundary failure that could fail a safety function or challenge the ability to reach and maintain safe shutdown is stated to be addressed by pre-determined HSS criteria, specifically Criteria 5, 6, 7, 8, and 11. Criterion 11 is credited for accounting for loss of functions that are not accounted for in Criteria 1 through 10, due to indirect effects (e.g., loss of power due to flooding).
For Questions 3 and 4, EPRI also credits Criteria 1 through 11 for addressing the concerns, and also states that the BWR Owner's Group and PWR Owner's Group have evaluated the standard plant Emergency Operating Procedures and no instances of any components were found to be the sole means for successful performance of actions required to mitigate an accident or transient, or for assuring long-term containment integrity, monitoring of post-accident conditions, or offsite emergency planning activities.
For Question 5, EPRI states, in part, [t]he proposed methodology requires all Class 1 SSCs be HSS. Class 1 components compose one of the key fission-product barriers. The response also explains that Criterion 9 addresses certain components that could lead to containment bypass and that Criterion 11 would identify components that would lead to a large early release, and potentially offsite radiological protective actions.
Questions 6 - 10 further address the DID philosophy. The NRC staff notes these questions are consistent with guidance on DID in Section 2.1.1.2 of RG 1.174, Revision 3. In its response, EPRI states no further evaluation is required when implementing the proposed methodology because the 10 CFR 50.69 categorization process does not change the design, design basis, or operation of plant components, including the reliance on programmatic activities or operator actions, independence of fission-product barriers, and level of system redundancy, independence and diversity. For Question 9, EPRI states, [c]ommon cause is a fundamental aspect of the PRA consequence evaluation methodology and therefore is taken into account.
The NRC staff notes that PRA models would account for common-cause failures (CCFs) of active components that are risk significant but clarifies that CCFs are not typically considered for passive components (i.e., multiple and simultaneous pipe ruptures are not expected), as discussed in the Vogtle 10 CFR 50.69 SE (NRC 2014a).
The NRC notes that the guidance in Section 9.2.2 of NEI 00-04, as endorsed by RG 1.201, Revision 1, also states the IDP should consider items that would address these same ten questions as part of its review of risk information and DID implications. (RG 1.201 clarifies that the IDP review of risk information should address both active and passive functions and SSCs.)
Based on the above discussion, the NRC staff finds that the EPRI TR 3002025288 methodology
23 adequately addresses these qualitative considerations. The NRC staff concludes that the requirements of 10 CFR 50.69(c)(1)(iii) to maintain DID would be met for passive components under a 50.69 categorization program implemented in accordance with EPRI TR 3002025288 and NEI 00-04, as endorsed by RG 1.201.
3.2.1.5 External Events Evaluation (TR Section 4.3)
As discussed in the response to RAI 2b, Section 4.3, of EPRI TR 300205288 states that preliminary HSS and LSS assignments shall be reviewed and adjusted to reflect the pressure boundary failures impact on the mitigation of external events. The method requires that the effect of external events on core damage and containment performance shall be evaluated for external events that can cause pressure boundary failure (e.g., seismic events) and events that do not affect the likelihood of pressure boundary failure but create demands that might cause pressure boundary failure and events (e.g., fires). EPRI TR 3002025288 explains that the purpose of this review is to confirm (and adjust as necessary) that the assignment of HSS is valid in the context of other hazards (e.g., fire, seismic, other hazards).
The NRC staff notes that these requirements are equivalent to those in Section I-3.1.2(f),
External Events Evaluation, of the ANO-2 RI-RRA methodology. The use of qualitative assessment for external hazards (not modeled in the PRA) is also consistent with the guidance for categorization of active components in NEI 00-04.
The response to RAI 2b also refers to Section 4.5 of EPRI TR 3002025288, which adds a requirement for the IDP to confirm that the assignment of HSS criteria is valid in the context of other hazards (fire, seismic, other hazards). Paragraph 50.69(b)(2)(ii) of 10 CFR contains requirements that a LAR to adopt 10 CFR 50.69 include descriptions of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown are adequate for the categorization of SSCs. Similarly, 10 CFR 50.69(c)(1)(ii) requires the SSC categorization process to determine the SSC functional importance using an integrated, systematic process that addresses initiating events (internal and external), SSCs, and plant operating modes, including those not modeled in the PRA.
Based on the above discussion, the NRC staff finds that the approach in Section 4.3 of EPRI TR 300205288 is consistent with existing precedents and NRC-endorsed guidance and is acceptable for assessing the impact of external hazards for passive categorization. Therefore, the NRC staff also finds that this approach is consistent with the requirements of 10 CFR 50.69(c)(1)(ii) to determine SSC functional importance using an integrated, systematic process that includes addressing external initiating events that are not modeled in the PRA.
3.2.1.6 Shutdown Risk (Response to RAI 4)
Section I-3.1.2(e), Shutdown Operation Evaluation, of the ANO-2 RI-RRA methodology includes requirements to review and adjust the consequence rank to reflect the pressure boundary failures impact on plant operation during shutdown. In practice, this assessment is qualitative in nature because licensees have not credited a shutdown PRA as part of their requests to adopt 10 CFR 50.69. The ANO-2 RI-RRA methodology provides major characteristics to be considered such as system operations and safety functions for different modes of operation but does not provide specific guidance on how to perform the assessment.
24 In response to RAI 4a, EPRI states that shutdown risk is evaluated consistent with NUMARC 91-06 (NUMARC 1991) for all stations, with a focus on protecting decay heat removal DID. The NRC staff notes that this is consistent with information provided in LARs to adopt 10 CFR 50.69 and that the use of NUMARC 91-06 is consistent with the guidance in NEI 00-04.
In the methodology of Chapter 4 of EPRI TR 3002025288, any pressure boundary failure that could fail a safety function is considered HSS. Per the specific criteria: Criterion 1 (reactor pressure boundary), Criterion 2 (applicable portions of the shutdown cooling pressure boundary function), Criterion 5 (loss of UHS), Criterion 6 (loss of ECCS), Criterion 7 (loss of secondary cooling in a PWR), Criterion 8 (loss of component cooling water (CCW) in a PWR), and Criterion 9 (heat exchangers that interface with reactor coolant system) apply. These systems are relied upon during shutdown conditions and also for decay heat removal and inventory control. Hence, the NRC staff finds that the passive categorization process in EPRI TR 3002025288 is consistent with existing precedents and sufficiently considers specific losses of key safety functions that could impact shutdown operations and no additional specific review is required for pressure boundary failures during shutdown conditions. The NRC staff also finds that this approach is consistent with the requirements of 10 CFR 50.69(c)(1)(ii) to determine SSC functional importance using an integrated, systematic process that includes addressing low power and shutdown operation modes that are not modeled in the PRA.
3.2.1.7 Integrated Decision Making Panel (IDP) Guidance (TR Section 4.5)
In response to RAI 2c regarding the role of the IDP in the passive pressure boundary categorization process, EPRI added Section 4.5 to EPRI TR 3002025288 to further describe the role of the IDP in passive pressure boundary categorization. The text states:
After the performance of the evaluations required by Sections 4.1 (Prerequisites),
4.2 (Determination of HSS Passive SSCs), 4.3 (External Events Evaluation), and 4.4 (Acceptably Small Increases to CDF and LERF), a preliminary (candidate)
HSS / LSS assignment of all safety-related and non-safety-related pressure-retaining components has been completed.
As required by Section 9.2 of NEI 00-04, the IDP is responsible for reviewing candidate HSS and LSS assignments and determining the final HSS and LSS assignment. Consistent with past practice any candidate HSS assignment (i.e., components meeting any one of the 11 criteria or determined to be HSS by a non-PRA external hazard evaluation) cannot be assigned LSS by the IDP. Per NEI 00-04, the IDP may determine a function/SSC has not been appropriately characterized and may be re-evaluated based on insights from the IDP. Also, NEI 00-04 allows for more detailed characterization of the SSC associated with a safety-significant function. This can be performed after the initial IDP, but the basis for that re-categorization must be considered and discussed in a follow up IDP session.
For application of the categorization methodology for pressure boundary components the IDP shall also confirm that all steps in the process have been followed.
The IDP shall ensure that the prerequisites cited in Section 4.1 are met.
25 The IDP shall confirm the assignment of HSS components (from the results of using Criteria 1 through 11) is appropriate.
The IDP shall confirm that the assignment of HSS criteria is valid in the context of other hazards (fire, seismic, other hazards).
In the response to RAI 2b, EPRI states that, from a practical implementation perspective, the passive/pressure boundary categorization will be presented to the IDP as a complete package (e.g., the full plant evaluation of the pressure retention function, that is all safety-related and non-safety-related pressure boundary components) for final categorization. After this categorization, as new systems are categorized - each systems results will include the active functions for IDP review and concurrence. The active functions will continue to be categorized consistent with guidance in NEI 00-04 (NEI 2005) as currently performed. This sequence assures that the IDP can assess the entire pressure boundary system at the beginning of the categorization process (IDP panel specific to categorization based on EPRI TR 3002025288) and assures that each subsequent system characterization also reflects the entire NEI 00-04 process.
3.2.2.
Risk-informed Evaluation The following sections summarize how the five key principles of risk-informed decisionmaking in RG 1.174, Revision 3, are met, in part, by implementation of a passive categorization process performed in accordance with EPRI TR 3002025288.
3.2.2.1 Key Principle 1: Licensing Bases Change Meets the Current Regulations As discussed in Section 3.2.2.4 of this SE, the use of Criterion 11 and Prerequisite 1 ensures that the method in Chapter 4 of EPRI TR 3002025288 considers results from the plant-specific PRA and that the PRA is peer reviewed and of sufficient quality level and detail to support the PRA categorization process, so that 10 CFR 50.69(c)(1)(i) would be met if this method was incorporated into a 10 CFR 50.69 program.
As discussed in Sections 3.2.1.2 and 3.2.1.4 through 3.2.1.7 of this SE, the EPRI TR 3002025288 method (i.e., Prerequisites, Criteria 1 through 12, External Events Evaluation and reliance on the IDP) adequately incorporates risk insights from the internal events PRA (with internal flooding), accounts for external events and shutdown risk, and uses an integrated, systematic process to evaluate the pressure boundary function such that 10 CFR 50.69(c)(ii) would be met if this method was incorporated into a 10 CFR 50.69 program.
As summarized in the next section, the method would ensure DID is maintained so that 10 CFR 50.69(c)(1)(iii) would be met if this method was incorporated into a 10 CFR 50.69 program.
As discussed in Section 3.2.1.4 of this SE, the NRC staff found that the use of Criterion 11 in conjunction with Prerequisites 1 and 2 provides reasonable confidence that for passive RISC-3 SSCs sufficient safety margins are maintained and any potential increase in CDF and LERF resulting from changes in treatment are small. Therefore, 10 CFR 50.69(c)(1)(iv) would be met if this method was incorporated into a 10 CFR 50.69 program.
26 As discussed in Section 3.2.1.3 of this SE, the passive categorization methodology would comply with the requirements in 10 CFR 50.69(c)(1)(v) to perform categorization for entire systems based on its definition of the pressure boundary function as its own system.
As discussed in Section 3.2.1.7 of this SE, the methodology relies on an IDP, in accordance with NEI 00-04, as endorsed by RG 1.201, Revision 1, therefore 10 CFR 50.69(c)(2) would be met if this method was incorporated into a 10 CFR 50.69 program.
Based on the above, the NRC staff finds that the method in Chapter 4 of EPRI TR 3002025288 is consistent with the requirements of 10 CFR 50.69(c)(1) and (2) and is suitable for use in applications to adopt or amend use of 10 CFR 50.69. Therefore, the NRC staff concludes that the first key principle for risk-informed decisionmaking identified in RG 1.174, Revision 3, would also be satisfied.
3.2.2.2 Key Principle 2: Licensing Basis Change is Consistent with the DID Philosophy The review in Section 3.2.1.4 of this SE, specifically for Criteria 1 through 11, as well as the additional DID and qualitative considerations, determined that the results of passive categorization using the method in Chapter 4 of EPRI TR 3002025288 would be appropriate because ASME Code Class 1 SSCs would remain HSS, components whose pressure boundary failures failed a key safety function would be categorized as HSS, IDP assurance that DID is maintained for both PRA-modeled and unmodeled SSCs, and the use of CCDP/CLERP risk metrics to designate components as HSS, such that no additional assessment related to DID was found to be required. Furthermore, HSS designations based on the method are expected to largely be consistent with results using the approved ANO-2 RI-RRA method and would therefore offer a similar level of DID. Additionally, an IDP is still required to consider DID as part of its review.
Therefore, the NRC staff concludes that a 10 CFR 50.69 program which adopts the passive categorization method in Chapter 4 of EPRI TR 3002025288 would be able to demonstrate, for passive components, that the second key principle for risk-informed decisionmaking is satisfied and the requirement of 10 CFR 50.69(c)(1)(iii) that DID be maintained is met.
3.2.2.3 Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins The third key risk-informed principle in RG 1.174 states that the licensing basis change should maintain sufficient safety margins. The engineering evaluations that will be conducted by a licensee under 10 CFR 50.69 for SSC categorization, using the method in Chapter 4 of EPRI TR 3002025288, in conjunction with NEI 00-04, as endorsed by RG 1.201, Revision 1, will assess the design function(s) and risk significance of SSCs to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring the categorization of the SSC does not adversely affect any assumptions or inputs to the safety analysis; or, if such inputs are affected, providing justification that sufficient safety margin will continue to exist.
The NRC staff notes that the design-basis functions of SSCs as described in a plants licensing basis, including the Updated Final Safety Analysis Report and technical specification bases, do not change and the safety margins described should continue to be met. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. On this
27 basis, the NRC staff concludes that safety margins are maintained by the proposed method, and the third key safety principle identified in RG 1.174, Revision 3, is satisfied.
3.2.2.4 Key Principle 4: Change in Risk is Consistent with the Safety Goals PRA Acceptability Pursuant to 10 CFR 50.69(c)(1)(i), the 10 CFR 50.69 categorization process must consider results and insights from a plant-specific PRA. The PRAs must be acceptable to support the categorization process and must be subjected to a peer review process. As discussed in Section 3.2.1.4 of this SE, the use of an internal events PRA (including internal flooding) that meets the requirements in Prerequisite 1 would be sufficient for application of this methodology as part of a 10 CFR 50.69 program. Additionally, the use of Criterion 11 addresses the requirement to consider results and insights from the PRA. Therefore, the NRC staff concludes that a licensee that adopts the passive categorization method in EPRI TR 3002025288 would be able to demonstrate that the requirements set forth in 10 CFR 50.69(c)(1)(i) are met.
The NRC staff notes that if the plant-specific PRA model(s) was previously found acceptable for use with the ANO-2 passive categorization method then the models would also be acceptable for use with methodology in EPRI TR 3002025288. However, if the plant had implementation items (e.g., changes to the PRA required to disposition peer review findings) related to passive categorization or the internal flooding PRA, then the status of these items should be provided as part of a LAR to adopt this topical.
Credit for Operator Actions and FLEX In response to RAI 3d, EPRI stated that human actions/recovery actions credited in the PRA must satisfy the requirements of the ASME/ANS PRA Standard (ASME 2009a). The technical element - Human Reliability Analysis (HRA) of Part 2 of the ASME/ANS PRA Standard outlines the requirements for human actions including that the action(s) must be proceduralized, must address plant-specific and scenario-specific influences on human performance, as well as consider the timing and availability of cues. Recovery actions shall only be modeled if the action is plausible and feasible.
Additionally, for internal flooding, the following additional supporting requirements apply, specifically:
IFQU-A5 (i.e., ensuring additional human failure events are in accordance with the human reliability requirements in Part 2) and IFQU-A6 (i.e., accounting for flood scenario-specific performance shaping factors such as additional workload and stress, cue availability, effect of flood on mitigation, timing and recovery actions, etc.).
Both ASME Code Case N-660 and ANO-2-R&R-004, Revision 1, credit operator actions required to mitigate flood impact by isolating valves typically from the control room. In ANO-2-R&R-004, PRA technical adequacy of modeling operator actions is addressed through NRC staff endorsement of the RI-ISI EPRI Report TR-112657, Revision B-A, which contains a requirement that licensees need to make a statement that their PRA is of sufficient quality.
Therefore, subsequent approvals to use the ANO-2 RI-RRA for passive categorization as part of 10 CFR 50.69 program were assessed against RG 1.200. (NRC 2015a). The NRC staff finds
28 that licensee adherence to the ASME/ANS standard supporting requirements for internal flooding events are sufficient to model potential operator actions for flood mitigation.
For HRA, in addition to the ASME/ANS PRA Standard requirements for FLEX, the NRC has issued a memorandum (NRC 2022b) on modeling of FLEX actions that need to be considered.
In the NRC 2022 memorandum (NRC 2022a), the NRC updated its assessment of NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, originally published in an NRC 2017 memorandum (NRC 2017c). With respect to FLEX actions, most licensee credited actions are fed through loss-of-offsite power events that go to station blackout (SBO) with late failures or direct SBO. Most pressure boundary initiators are generally mapped to transients (e.g., reactor trip or loss of cooling system initiators). For pressure boundary failures, multiple non-pressure-retaining related failures would need to occur before typical FLEX actions would be credited (likely near the truncation of the internal flooding model). Therefore, the NRC staff finds that the uncertainties associated with the PRA modeling of FLEX equipment and operator actions would not be expected to impact passive categorization using the methodology in Chapter 4 of EPRI TR 3002025288.
Acceptably Small Increases to CDF and LERF (TR Section 4.4)
As discussed in Section 3.2.1.4 of this SE, the use of Criterion 11 to identify significant outliers in terms of risk and consequences is only acceptable when applied to passive components identified as preliminary LSS via Criteria 1 through 10. The risk criteria are consistent with industry precedents for different types of applications and are intentionally less conservative than the risk metrics provided in the ANO-2 RI-RRA method. It is noted that the risk criteria in ANO-2 RI-RRA do not explicitly account for uncertainty. The following sections discuss consideration of uncertainty and EPRI TR 3002025288 requirements for a cumulative sensitivity study.
Consideration of Uncertainty To account for uncertainty in the future reliability and incidence rate of passive failure rates, Section 4.4 of EPRI TR 3002025288 and the response to RAI 3a, required that pressure boundary components that are modeled in the internal events or internal flooding PRA that have been determined to be LSS shall have their failure rates (such as pipe break frequency) increased by a factor of 3 and their cumulative CDF and LERF re-quantified. The results of this sensitivity study are then required to be compared to the quantitative acceptance guidelines of RG 1.174 to determine if estimated CDF/LERF values would exceed the guidelines. Section 4.3 of EPRI TR 3002025288 further states: Any pressure boundary component as modeled in the
[internal events] PRA or [internal flooding] PRA that exceeds the acceptance guidelines of RG 1.174 shall be candidate highsafetysignificant, subject to IDP concurrence.
The factor of 3 is traditionally used for sensitivity analysis of active systems per NEI 00-04 (NEI 2005); where it is suggested that a factor of 3-5 be applied to identify if changes in the reliability of systems would have significant impacts on safety (e.g., conduct sensitivity studies, to estimate where potential changes in reliabilities may cause undue changes to risk conclusions). The factor of approximately 3 is used when assessing sensitivity to uncertainty as it is an approximation of the likely tail of a distribution for active systems. By applying this factor, an analysis accounts for uncertainty whether existing data represents the true average versus a nonconservative measure of that average due to sampling or estimation biases. When
29 altering the general approach (e.g., changing from HSS treatment to LSS treatment for passive systems), it is unclear why it is reasonable to assess the future distribution (LSS treatment) as matching the prior distribution (HSS treatment). Consequently, there is no current technical basis provided to estimate the change in failure frequency when going from an ASME Code inspection and repair-replacement program to an owners program without delineated minimum requirements.
Additionally, as discussed in Vogtle 10 CFR 50.69 SE (NRC 2014a), the plants performance monitoring process is intended to monitor component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study. Passive components have much lower historical failure rates. Consequently, it is possible for one or two failure events to cause a significant change in the failure frequency for a specific type of piping such that the factor of 3 would be exceeded for that subset of components and the intent would not be met. It is noted that a sensitivity study is not required as part of the ANO-2 RI-RRA methodology, although that methodology does have strict thresholds for CCDP/CLERP.
However, as discussed in detail in the applicants response to RAI 1a, it is expected that applicants will continue to prefer like-for-like repairs or replacement for RISC-3 SSCs. Based on this, the NRC staff understands that generally repairs and replacements will continue to be like-for-like as described in the RAI response. Like-for-like repairs are described as industrial grade physically and functionally the same as previously supplied for safety-related use. This was further explained during the audit related meetings as ordering the same part, but without many aspects otherwise required for RISC-1 applications (such as 10 CFR Part 50, Appendix B, required documentation).
The NRC staff concur that like-for-like repairs and replacements are sufficiently similar to RISC-1 practice that using an uncertainty factor of 3 provides a reasonable sensitivity study for the statistical population of RISC-3 SSCs. This is because the factor of 3 provides an estimate of bounding reliability for a particular distribution of components given that like-to-like repair or replacements are expected to result in largely similarly reliabilities (e.g., be statistically similar to prior non-RISC-3 operating experience reliability).
For the sub-population of repair or replacements not conducted as like-for-like, the NRC staff reviewed the response to RAI 1a (EPRI 2025a), and the provided description of a pressurized water reactor owners group (PWROG) document titled Supply Chain, Procurement Engineering, and Design Engineering Roadmap for Procurement of RISC-3 Items. Based on this information, the NRC staff finds that the likelihood of substantive reliability deviations beyond using a factor of 3 for sensitivity study of RISC-3 systems is acceptably small because the subject population potentially falling beyond the factor of 3 is likely to be small.
The NRC staff does not, at this time, have information to conclude that like-for-like repair or replacement is no longer the predominant approach, or that general industry practice for non-like-for-like activities is sufficiently common and impactful as to require a higher factor for sensitivity studies. The remaining plant-specific considerations managed through application of Criterion 11 are a relatively small portion of the total plant risk because the majority of DID and risk-significant SSCs are categorized as HSS by Criteria 1 through 10. Consequently, the NRC staff consider the management of the residual risks through Criterion 11 adequate because Criteria 1 through 10 identify the majority of systems that should be designated HSS per
30 risk-informed analysis and applying the factor of 3 as a sensitivity study for the remaining plant-specific SSCs is sufficient for this methodology as discussed above. However, the NRC staff is not generically approving the use of a factor of 3 for passive categorization applications, particularly for changes for which there is limited data available.
As discussed in Section 3.2.1.4 of this SE, the updated Criterion 11 accounts for CCDP/CLERP, as well as CDF/LERF, so that components with values of CCDP/CLERP >1x10-2/1x10-3 would still be identified as preliminary HSS. When also considering the CDF/LERF metrics of 1x10 6/yr / 1x10-7/yr, this means that the maximum allowable rupture frequency would be 1x10 4/yr. Even for repairs that involve material changes for which there is limited failure data, it is unlikely that an increase in failure frequency of multiple orders of magnitude would occur such that there would be an immediate concern with respect to public health and safety. Additionally, the design change control process, periodic inspection and testing required by 10 CFR 50.69(d)(2)(i), and corrective action program would also be expected to limit the potential for a significant increase in failure rates.
In the response to RAI 3a, EPRI states, in part: The analyst must review key assumptions and sources of model uncertainty in the context of this application. This statement was confirmed in correspondence dated June 11, 2025 (EPRI 2025b). As discussed in Section 3.2.1.2 of this SE, Prerequisite 1, requires that key assumptions and sources of model uncertainty be reviewed in context of the application and that any non-conservatisms or the use of methods not commonly accepted for risk-informed applications be reviewed to determine their impact. As discussed in the response to RAI 3a, only those assumptions or sources of uncertainty that could significantly impact the categorization risk calculations (i.e., could change a categorization outcome) would be considered key for this application. If a licensee were to later perform a design change that could significantly impact the application (i.e., could change the 50.69 risk categorization for a component), this would be expected to be identified by the licensee as part of their PRA configuration control and feedback and adjustment processes, and any new key source of uncertainty would be addressed, as needed. (Prerequisite 4 addresses PRA configuration control requirements and 10 CFR 50.69(e) includes requirements for feedback and process adjustment.)
Finally, the NRC staff performed a review of risk results in recent submittals (including 10 CFR 50.69 submittals) as well as internal confirmatory analysis using SPAR-DASH results for several plants and determined that even if flood risk was tripled (which is conservative as some passive SSCs would remain RISC-1) the majority of plants would not be expected to exceed the risk acceptance guidelines in RG 1.174. Therefore, the cumulative sensitivity study is expected to be of limited value and would not be expected to cause a significant number of SSCs to be changed from preliminarily LSS to preliminarily HSS. Additionally, it is noted that categorization of an SSC as HSS due to the cumulative sensitivity study is subject to IDP concurrence and can be changed back to LSS.
Based on the above, the NRC staff concludes that the acceptability of the methodology is not dependent on performing sensitivity studies using a factor of 3. The use of Criterion 11, which considers both CDF/LERF and CCDP/CLERP risk metrics, in combination with Prerequisite 1, which requires a technically adequate PRA model and for a risk analyst to review key assumptions and sources of uncertainty in the context of the application and review non-conservatisms or the use of methods not commonly accepted to determine their impact, and Prerequisite 2, which includes requirements related to programs for ensuring integrity
31 management, is sufficient to support findings under 10 CFR 50.69(c)(1)(iv). Specifically, the NRC staff finds that the method in Chapter 4 of EPRI TR 3002025288 includes evaluations that provide reasonable confidence that for passive RISC-3 SSCs sufficient safety margins are maintained and any potential increase in CDF and LERF resulting from changes in treatment are small.
The NRC staff also concludes that any changes in CDF and LERF resulting from use of the method in EPRI TR 300205288 as part of a 10 CFR 50.69 categorization process are expected to be acceptably small consistent with NRC policies for risk-informed approaches and the fourth key principle of risk-informed decisionmaking is satisfied.
3.2.2.5 Key Principle 5: Monitor the Impact of the Proposed Change Key principle 5 requires that the impact of the proposed licensing basis change should be monitored using performance measurement strategies. EPRI TR 3002025288, Prerequisite 4, provides guidance for the asbuilt asoperated plant, that the PRA must have a configuration control program, update process, and periodically update to the PRA as described below:
A PRA Configuration Control Program shall be in place in accordance with the ASME/ANS PRA Standard that maintains the PRA, and any supplementary analysis, to reflect the asbuilt/asoperated plant.
A feedback and adjustment process shall be in place to review changes to the plant, operational practices, applicable plant and industry operational experience, and as appropriate, update the PRA model and if necessary, the categorization and treatment processes.
This review shall meet the required update periodicity requirements for input information, such as the PRA update requirements for the underlying information used as input in the methodology implementation.
Licensees that have adopted 10 CFR 50.69 currently address these requirements for PRA maintenance/configuration control, and feedback and process adjustment in accordance with NEI 00-04, as endorsed by RG 1.201, Revision 1. Meeting the requirements in Prerequisite 4 would be consistent with 10 CFR 50.69(c)(ii) and (e).
In addition, the operational programs described in Section 4.1.2 of EPRI TR 3002025288 provide substantial performance monitoring of subject SSCs susceptible to the listed degradation mechanisms. In the response to RAI 6c, EPRI also updated Section 4.1.2 of TR 3002025288 to clarify that the requirements of 10 CFR 50.69(e) are also applicable to the integrity management programs required by Prerequisite 2.
Considering the above, the NRC staff has determined that the proposed change satisfies the fifth key principle for risk-informed decisionmaking identified in RG 1.174 and that the methodology would continue to be acceptable for use during the performance monitoring phase.
32
- 4.
LIMITATIONS AND CONDITIONS The NRC staff did not identify the need to include or issue any limitations or conditions for the use of the approved version of EPRI TR 3002025288.
- 5.
CONCLUSION Based on information provided in EPRI TR 3002025288, as supplemented, the NRC staff finds that SSC categorizations made using the methodology in Chapter 4 of EPRI TR 3002025288 would be consistent with the principles of risk-informed decisionmaking in RG 1.174, Revision 3.
The NRC staff concludes that categorization of passive, pressure-retaining components in accordance with the methodology in Chapter 4 of the approved version of EPRI TR 3002025288 is acceptable for use as part of a licensees 10 CFR 50.69 categorization process.
- 6.
REFERENCES Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities 10 CFR Part 50, Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors" 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 10 CFR Part 50, Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979" 10 CFR Part 54, Requirements for renewal of operating licenses for nuclear power plants 10 CFR Part 100, "Reactor site criteria" 10 CFR Part 100, Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants" 69 FR 68008, November 15, 2004, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors; Federal Register, Nuclear Regulatory Commission.
ASME, 2005a, "Case N-660b, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities," March 25, 2005 (ADAMS Accession No. ML051880307)
ASME, 2009a, "American Society of Mechanical Engineers / American Nuclear Society,
'Addenda to ASME/ANS RA-S-2008, Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009,"
New York, NY, February 2009 Entergy Nuclear Operations, Inc., Palisades Nuclear Plant, 2014a, Letter from Jeffery A. Hardy, Entergy, to NRC, "Relief Request Number RR 4-20 Proposed Alternative, Use of Alternate ASME Code Case N-716," August 14, 2014 (ADAMS Accession No. ML14226A6180)
33 EPRI, 1994, TR-103403, "Service Water System Corrosion and Deposition Sourcebook,"
March 1, 1994 (Publicly available free from EPRI.com)
EPRI, 2000a, Letter from Jeffrey Mittman to NRC, "EPRI Topical Report TR-112657 Revision B-A, Revised Risk-Informed Inservice Inspection Procedure. Reference Project #669,"
February 10, 2000 (ADAMS Accession No. ML013470102)
EPRI, 2011, Report 1022945, Risk-Informed Repair/Replacement Methodology, October 28, 2011 (not publicly available)
EPRI, 2023a, Letter from Fernando Ferrante, EPRI, to NRC, "Request for NRC Review of
'Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components,'
EPRI Technical Report 3002025288, June 2023," August 17, 2023 (ADAMS Accession No. ML23234A266).
EPRI, 2023b, Letter from Fernando Ferrante, EPRI, to NRC, "Electric Power Research Institute
- Submittal of Supplemental Information Needed to Complete Acceptance Review of technical Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components," November 30, 2023 (ADAMS Accession No. ML23334A210)
EPRI, 2023c, "2023 Technical Report 3002025288 - Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components, June 30, 2023 (ADAMS Accession No. ML23234A268)
EPRI, 2024a, Letter from Fernando Ferrante, EPRI, to NRC, "Electric Power Research Institute
- Submittal of Supplemental Information to Complete Acceptance Review of Technical Report 3002025288, Enhanced Risk-Informed Categorization Methodology For Pressure Boundary Components," June 28, 2024 (ADAMS Accession No. ML24180A016)
EPRI, 2025a, Letter from Fernando Ferrante, EPRI, to NRC, "Electric Power Research Institute
- Submittal of Responses to the US NRC's Request for Additional Information (January 2025) on EPRIs Technical Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components," April 30, 2025, (ADAMS Accession No. ML25121A201)
EPRI, 2025b, E-mail from Ashley Lindeman, EPRI, to Lois James, NRC, RE: Proposed Limitations and Conditions to be Contained in the Safety Evaluation for EPRI Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components (EPID No. L-2023-NTR-0008), June 11, 2025 (ADAMS Accession No. ML25163A050)
NEI, 2005, NEI 00-04 Technical Report, "10 CFR 50.69 SSC Categorization Guideline,"
Revision 0, July 31, 2005 (ADAMS Accession No. ML052910035)
34 NEI, 2016a, Letter from Anthony R. Pietrangelo, EPRI, to William M. Dean, NRC, "Transmittal of NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Revision 0, For Information Only," August 26, 2016 (ADAMS Accession No. ML16286A297)
NEI, 2021, NEI 17-06 Technical Report, Revision 1, Guidance on Using IEC 61508 SIL Certification to Support the Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Related Applications, December 3, 2021 (ADAMS Accession No. ML21337A380)
NRC, 2004a, "SECY-04-0109 - Response to Public Comments to Final Rule 10 CFR 50.69,"
November 10, 2004 (ADAMS Accession No. ML042990011)
NRC, RG 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, Revision 1, May 2006 (ADAMS Accession No. ML061090627)
NRC, 2009a, Letter from Michael Markley, NRC to Arkansas Nuclear One-Entergy Operations, Inc., 2009, "Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO2-R&R-004, Revision 1, to Use Risk-Informed Safety Classification and Treatment Repair/Replacement Activities in Class 2 & 3 Moderate Energy Systems (TAC No. MD5250),"
April 22, 2009 (ADAMS Accession No. ML090930246)
NRC, 2009b, NRC RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities," Revision 2, March 1, 2009 (ADAMS Accession No. ML090410014)
NRC, 2010a, NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants - Final Report," December 2010 Revision 2 (ADAMS Accession No. ML103490036)
NRC, 2012a, Letter from Menna Khanna, NRC, to Kevin Walsh, NextEra Energy Seabrook, LLC, "Seabrook Station, Unit 1 -Relief for Alternative 3AR-1, Use of a Risk-Informed, Safety-Based Inservice Inspection Program (TAC No. ME7569)," June 21, 2012 (ADAMS Accession No. ML121320552)
NRC, 2014a, Letter from Robert Martin, NRC, to C. R. Pierce, Southern Nuclear Operating Company, Inc., "Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments (TAC Nos. ME9472 and ME9473)," December 17, 2014 (ADAMS Accession No. ML14237A034)
NRC, 2017a, NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking - Final Report," Revision 1, March 2017(ADAMS Accession No. ML17062A466)
35 NRC, 2017b, NUREG-2192, "Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants - Final Report," July 2017 (ADAMS Accession No. ML17188A158)
NRC, 2017c, Memorandum from Mehdi Reisi-Fard, NRC, to Joseph G. Giitter, NRR, "Assessment of the NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis," May 30, 2017 (ADAMS Accession No. ML17031A269)
NRC, 2020, NRC RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3, December 2020 (ADAMS Accession No. ML20238B871)
NRC, 2021a, Safety Evaluation for "Arkansas Nuclear One, Units 1 and 2 - Approval of Request for Alternative From Certain Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (EPID L-2020-LLR-0076), May 19, 2021 (ADAMS Accession No. ML21118B039)
NRC, 2022a, NRC RG 1.250, Dedication of Commercial-Grade Digital Instrumentation and Control Items for Use in Nuclear Power Plants, Revision 0, October 2022 (ADAMS Accession No. ML22153A408)
NRC, 2022b, Memorandum from Antonios M. Zoulis, NRC, to Michael X. Franovich, NRC, "Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments," May 6, 2022 (ADAMS Accession No. ML22014A084)
NRC, 2023a, letter from Bo M. Pham, NRC, to Steven M. Snider, Duke Energy Carolinas, LLC, "Oconee Nuclear Station, Units 1, 2, and 3 - RE: Alternative to Use RR-22-0174, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1 (EPID L-2022-LLR-0060)," December 13, 2023 (ADAMS Accession No. ML23262A967)
NRC, 2024a, E-mail from Lois James, NRC, to Ferando Ferrante, EPRI, "NRC Completeness Determination Form 898 for EPRI Technical Report 3002025288 Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components (Risk-Informed CMPBC)
(EPID L-2023-TOP-0045)," July 11, 2024 (ADAMS Accession No. ML23352A054)
NRC, 2024b, E-mail from Lois James, NRC, to Ferando Ferrante, EPRI, "Electric Power Research Institute Regulatory Audit in Support of Review of Technical Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components, Revision 0 (EPID L-2023-TOP-0045)," August 30, 2024 (ADAMS Accession No. ML24241A160)
NRC, 2024c, RG 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 21, March 2024 (ADAMS Accession No. ML23291A003)
36 NRC, 2024d, Letter from Jennivine K. Rankin, NRC, to Ronald Gaston, Entergy Services, LLC, "Grand Gulf Nuclear Station, Unit 1; River Bend Station, Unit 1; and Waterford Steam Electric Station, Unit 3 - Authorization of Alternative to Use EN-RR-01 Concerning Proposed Alternative to Adopt Code Case N-752 (EPID L-2022-LLR-0054)," May 30, 2024 (ADAMS Accession No. ML24151A00-04236)
NRC, 2024d, Letter from Jamie Pelton, NRC, to Bob Coffey, NextEra Energy Seabrook, LLC, "NextEra Fleet - Proposed Alternative FRR 23-01 to use ASME Code Case N-752-1, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems Section X1, Division 1 (EPID L-2023-LLR-0009)," June 12, 2024 (ADAMS Accession No. ML24149A305)
NRC, 2025a, E-mail from Lois James, NRC, to Fernando Ferrante, EPRI, "Final Request for Additional Information - EPRI Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components (EPID No. L-2023-NTR-0008), January 13, 2025 (ADAMS Accession No. ML24352A469)
NRC, 2025b, E-mail from Lois James, NRC, to Fernando Ferrante, EPRI, "Regulatory Audit Report in Support of Review of Electric Power Research Institute Technical Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components, Revision 0 (EPID L-2023-TOP-0045)," June 12, 2025 (ADAMS Accession No. ML25147A118)
NRC, 2025c, Letter from Scott P. Wall, NRC, to Kelly J. Ferneau, Indiana Michigan Power Company, " Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 Issuance of Amendment Nos. 365 and 346 Regarding Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2024-LLA-0025)," April 9, 2025 (ADAMS Accession No. ML25071A389)
NRC, 2025d, Letter from Aida Rivera-Varona, NRC, to Jamie M. Coleman, Southern Nuclear Operating Co., Inc., "Joseph M. Farley Nuclear Plant, Units 1 and 2; Edwin I. Hatch Nuclear Plant, Units 1 and 2; and Vogtle Electric Generating Plant, Units 1 and 2 - Proposed Alternative to use ASME Code Case N-752 (EPID L-2024-LLR-0044)," March 27, 2025 (ADAMS Accession No. ML24198A062)
Nuclear Management and Resources Council, Inc. (NUMARC), "Guidelines for Industry Actions to Assess Shutdown Management," December 1991 (ADAMS Accession No. ML14365A203)
- 7.
ABBREVIATIONS ANO Arkansas Nuclear One ASME American Society of Mechanical Engineers BWR boiling water reactors
37 CCDP conditional core damage probabilities CCF common-cause failures CCW component cooling water CDF core damage frequency CFR Code of Federal Regulation CLERP conditional large early release probabilities DID defense-in-depth DRA Division of Risk Assessment EPRI Electric Power Research Institute FV Fussell-Vesely HRA human reliability analysis HSS high safety significant IDP integrated decisionmaking panel ISI Informed inservice inspection LAR license amendment request LERF large early release frequency LOCA loss-of-coolant accident LSS low safety significant NEI Nuclear Energy Institute NPHP Division of New and Renewed NRC Nuclear Regulatory Commission PRA Probabilistic risk assessment PWR pressurized water reactors RAI requests for additional information RCS reactor coolant system RISC Risk-Informed Safety Class SBO station blackout SE safety evaluation SSC structures, systems, and components TR technical report