ML24198A062

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Proposed Alternative to Use ASME Code Case N-752
ML24198A062
Person / Time
Site: Hatch, Vogtle, Farley  
Issue date: 03/27/2025
From: Rivera-Verona A
Plant Licensing Branch II
To: Coleman J
Southern Nuclear Operating Co
Lamb J
Shared Package
ML24198A073 List:
References
EPID L-2024-LLR-0044
Download: ML24198A062 (22)
v • d • e


Text

March 27, 2025 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

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 - AUTHORIZATION OF PROPOSED ALTERNATIVE NL-24-0143 TO USE ASME CODE CASE N-752, RISK-INFORMED CATEGORIZATION AND TREATMENT FOR REPAIR/REPLACEMENT ACTIVITIES IN CLASS 2 AND 3 SYSTEMS, SECTION X1, DIVISION 1 (EPID L-2024-LLR-0044)

Dear Jamie Coleman:

By letter dated June 27, 2024 (Agencywide Documents Access and Management System Accession No. ML24179A334), as supplemented by letter dated November 15, 2024 (ML24320A129), Southern Nuclear Operating Company (SNC, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for authorization to use proposed alternative NL-24-0143 to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, requirements at Joseph M. Farley Nuclear Plant (Farley),

Units 1 and 2; Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2; and Vogtle Electric Generating Plant (Vogtle), Units 1 and 2. Specifically, the licensee requested to use Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1, in lieu of certain requirements in ASME Section XI, Sub-Paragraphs IWA-1320, IWA-1400, IWA-4000, IWA-6211, IWA-6220, and IWA-6350.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(z)(1), SNC requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety. ASME Code Case N-752 has not been approved by the NRC staff or incorporated by reference for generic use. Therefore, the NRC staff reviewed the licensees request as plant-specific requests for Farley, Units 1 and 2, Hatch, Units 1 and 2, and Vogtle, Units 1 and 2.

The NRC staff has reviewed the proposed alternative, and, as set forth in the enclosed safety evaluation, the NRC staff determines that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the proposed alternative for the remainder of the fifth inservice inspection interval at Farley, Units 1 and 2, the remainder of the fifth inservice inspection interval at Hatch, Units 1 and 2, and the remainder of the fourth inservice inspection interval at Vogtle, Units 1 and 2.

J. Coleman All other ASME Code,Section XI, requirements for which an alternative was not specifically requested and authorized in this alternative remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

In its request, SNC mentioned in several places that risk-informed categorization process described in Code Case N-752 is consistent with the risk-informed categorization process described in 10 CFR 50.69. For the licensees facilities that have been approved to use a risk-informed categorization process under 10 CFR 50.69, this authorization does not change any of the licensees obligations with its approved programs for implementing 10 CFR 50.69.

If you have any questions, please contact the SNC Fleet Senior Project Manager John G. Lamb at 301-415-3100 or John.Lamb@nrc.gov.

Sincerely, Aida Rivera-Varona, Deputy Director Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-321, 50-366, 50-348, 50-364, 50-424, and 50-425

Enclosure:

Safety Evaluation cc: Listserv AIDA RIVERA-VARONA Digitally signed by AIDA RIVERA-VARONA Date: 2025.03.27 17:29:39 -04'00'

Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED ALTERNATIVE TO USE ASME CODE CASE N-752, RISKINFORMED CATEGORIZATION AND TREATMENT FOR REPAIR/REPLACEMENT ACTIVITIES IN CLASS 2 AND 3 SYSTEMS, SECTION XI, DIVISION 1 JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 EDWIN I. HATCH NUCLEAR, UNITS 1 AND 2 VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SOUTHERN NUCLEAR OPERATING COMPANY DOCKET NOS. 50-321, 50-366, 50-348, 50-364, 50-424, AND 50-425

1.0 INTRODUCTION

By letter dated June 27, 2024 (Agencywide Documents Access and Management System Accession No. ML24179A334), as supplemented by letter dated November 15, 2024 (ML24320A129), Southern Nuclear Operating Company (SNC, the licensee) requested the use of an alternative to certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to use Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems, Section X1, Division 1 in lieu of certain requirements in ASME Section XI, Sub-Paragraphs IWA 1320, IWA 1400, IWA-4000, IWA-6211, IWA-6220, and IWA-6350 at Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2; Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2; and Vogtle Electric Generating Plant (Vogtle), Units 1 and 2.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),

SNC requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety. ASME Code Case N-752 has not been approved by the U.S. Nuclear Regulatory Commission (NRC) staff or incorporated by reference for generic use.

Therefore, the NRC staff reviewed the licensees request as plant-specific requests for Farley, Units 1 and 2, Hatch, Units 1 and 2, and Vogtle, Units 1 and 2.

2.0 REGULATORY EVALUATION

2.1 Regulations The following requirements are applicable to this request:

Section 50.55a(g)(4), Inservice inspection standards requirement for operating plants, of 10 CFR

Section 50.55a(z)(1), Acceptable level of quality and safety, of 10 CFR

Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants.

2.2 Regulatory Guidance The NRC staff used the following guidance in the evaluation of this request:

Regulatory Guide 1.178, Revision 2, Plant-Specific, Risk-Informed Decisionmaking for Inservice Inspections of Piping, April 2021 (ML21036A105)

Regulatory Guide 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, May 2011 (ML100910006)

Regulatory Guide 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256)

Regulatory Guide 1.177, Revision 2, Plant-Specific Risk-Informed Decisionmaking:

Technical Specifications, January 2021 (ML20164A034)

Regulatory Guide 1.200, Revision 1, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, January 2007 (ML070240001)

Regulatory Guide 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009 (ML090410014)

Regulatory Guide 1.200, Revision 3, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, December 2020 (ML20238B871)

3.0 TECHNICAL EVALUATION

3.1 The Licensees Proposed Alternative The applicable ASME Code editions and addenda for the applicable inservice inspection (ISI) intervals are specified in the table below for each plant.

Plant ISI Interval ASME Section XI Code of Record Edition/Addenda Interval Start Interval End Farley, Units 1 and 2 5th 2007/2008 12/1/2017 11/30/2027 Hatch, Units 1 and 2 5th 2007/2008 1/1/2016 12/31/2025 Vogtle, Units 1 and 2 4th 2007/2008 5/31/2017 5/30/2027

3.2 ASME Code Components Affected

In its letter dated June 27, 2024, SNC stated:

This request applies to ASME Class 2 and 3 items or components except the following:

Class CC and MC items.

Piping within the break exclusion region [> Nominal Pipe Size (NPS) 4 (DN 100)] for high energy piping systems as defined by the Owner.

That portion of the Class 2 feedwater system [> NPS 4 (DN 100)] of pressurized water reactors (PWRs) from the steam generator (SG),

including the SG, to the outer containment isolation valve.

a.

NUREG-0800, Section 3.6.2 provides a method for defining this scope of piping.

3.3 Applicable Code Requirements In its letter dated June 27, 2024, SNC stated:

ASME Code,Section XI, Subsection IWA provides the requirements for repair/replacement activities including the following:

IWA-1320 specifies group classification criteria for applying the rules of ASME Section XI to various Code Classes of components. For example, the rules in IWC apply to items classified as ASME Class 2 and the rules in IWD apply to items classified as ASME Class 3.

IWA-1400(f) requires Owners to possess or obtain an arrangement with an Authorized Inspection Agency (AIA).

IWA-1400(j) requires Owners to perform repair/replacement activities in accordance with written programs and plans.

IWA-1400(n) requires Owners to maintain documentation of a Quality Assurance Program in accordance with 10 CFR 50 or ASME NQA-1 [Quality Assurance Requirements for Nuclear Facility Applications], Parts II and III.

IWA-4000 specifies requirements for performing ASME Section XI repair/replacement activities on pressure-retaining items or their supports.

IWA-6211(d) and (e), specify Owner reporting responsibilities such as preparing Form NIS-2, Owners Repair/Replacement Certification Record.

IWA-6220 repeats the IWA-4150 requirements that a Repair/Replacement Plan be prepared for all repair/replacement activities, requires Form NIS-2 be completed, provides the required timing for completion of Form NIS-2, identifies certification requirements for Form NIS-2, and includes the requirement for maintaining an index of Repair/Replacement Plans.

IWA-6350 specifies that the following ASME Section XI repair/replacement activity records must be retained by the Owner: evaluations required by IWA-4160 and IWA-4311, Repair/Replacement Programs and Plans, reconciliation documentation, and NIS-2 Forms.

d.

Code Case N-752 is based on the 2017 Edition of ASME Section XI while SNCs Code of record for HNP [Hatch, Units 1 and 2], FNP [Farley, Units 1 and 2], and VEGP [Vogtle, Units 1 and 2], is the 2007 Edition/2008 Addenda, except as noted in Section 2 of this request. Below is a cross reference for affected code paragraphs:

IWA-1400(g), (k), and (o) in the 2017 Edition are IWA-1400(f), (j),

and (n) in the 2007 Edition/2008 Addenda.

IWA-6211(f) and IWA-6212 in the 2017 Edition do not exist in or apply to the 2007 Edition/2008 Addenda.

3.4 Proposed Alternative In its letter dated June 27, 2024, SNC stated:

Pursuant to 10 CFR 50.55a(z)(1), SNC proposes to implement ASME Code Case N-752, without exception, as an alternative to the ASME Code requirements specified in Section 3. ASME Code Case N-752 provides a process for determining the risk-informed categorization and treatment requirements for Class 2 and 3 pressure retaining items or the associated supports as defined in Section 1. This requested implementation includes the categorization of passive

[safety-related structures, systems, or components] SSCs (e.g., piping) and implementation of alternative special treatment activities limited to the repair/replacement activities for Class 2 and 3 pressure retaining items or their associated supports. For components that have both active and passive functions, only the passive function will be categorized. The alternative treatments associated with ASME Code Case N-752 will not be applied to the parts/components associated with the active function. Code Case N-752 may be applied on a system basis or on individual items within selected systems. Code Case N-752 does not apply to Class 1 items.

The use of this proposed alternative is requested on the basis that requirements in Code Case N-752 will provide an acceptable level of quality and safety.

3.5

NRC Staff Evaluation

The NRC independently evaluated SNCs request to determine if the proposed alternative met an acceptable level of quality and safety. The NRC staff reviewed the proposed alternative as a risk-informed request because the proposed alternative includes the use of a risk-informed process described in Appendix I of Code Case N-752-1. In evaluating the SNCs proposed alternative, the NRC staff considered the past precedent of previous NRC plant-specific approvals related to risk-informed treatment of SSCs for nuclear power plants. Specifically, the NRC staff authorized the licensee for the Arkansas Nuclear One, Units 1 and 2 (ANO) to utilize alternative ANO2-R&R 004, Revision 1, for determining the risk-informed categorization and for implementing alternative treatment for repair/replacement activities on moderate and high energy Class 2 and 3 items at ANO-2. By letter dated April 22, 2009, the NRC staff authorized the alternative (ML090930246). The NRC did not endorse the ANO plant-specific method for generic use although many 10 CFR 50.69 programs have adopted the ANO precedent for plant-specific use in the categorization and treatment of passive SSCs.

3.5.1 Probabilistic Risk Assessment Technical Acceptability The proposed plant-specific approach for SNC utilizes industry experience gained through the ANO precedents and utilizes the risk-informed categorization process in Appendix I of Code Case N-752 for ASME Class 2 and 3 systems. The process requires confirmation of the technical adequacy of the probabilistic risk assessment (PRA) model for its risk-informed inservice inspection (RI-ISI) program to confirm the applicability for categorization, including verification of assumptions on equipment reliability. The alternative authorized for ANO2-R&R-004, Revision 1 for ANO, Unit 2 (ML071150108), demonstrated adequate PRA technical requirements, as outlined in the NRC staffs safety evaluation dated April 22, 2009 (ML090930246) and has been used by numerous nuclear power plants for risk-informed categorization and treatment of Class 2 and 3 systems.

The NRC staff approved the following amendments:

Amendment Nos. 173 and 155 were issued for Vogtle, Units 1 and 2, respectively, to allow the implementation of the provisions of 10 CFR 50.69, December 17, 2014 (ML14237A034).

Amendment Nos. 188 and 171 were issued for Vogtle, Units 1 and 2, respectively, to modify the TS to permit use of Risk-Informed Completion Times (RICTs) in accordance with Topical Report Nuclear Energy Institute (NEI) 06-09, Risk-Informed Technical Specification Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A. August 8, 2017 (ML15127A669).

Amendment Nos. 233 and 230 were issued for Farley, Units 1 and 2, respectively, to allow the implementation of the provisions of 10 CFR 50.69, and adds a license condition that identifies action items that need to be completed prior to implementing 10 CFR 50.69 at Farley, Units 1 and 2 and identifies possible changes to the categorization process that would require prior NRC approval, June 30, 2021 (ML21137A247).

Amendment Nos. 233 and 230 were issued for Farley, Units 1 and 2, respectively, to modify the TS to permit use of RICTs in accordance with Topical Report NEI 06-09, August 23, 2019 (ML19175A243).

Amendment Nos. 305 and 250 were issued for Hatch, Units 1 and 2, respectively, to allow the implementation of the provisions of 10 CFR 50.69, June 26, 2020 (ML20077J704).

Amendment Nos. 319 and 264 were issued for Hatch, Units 1 and 2, respectively, to permit the use of RICTs for actions to be taken when limiting conditions for operation are not met, December 22, 2022 (ML22297A146). The changes were based on Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.

Amendment Nos. 326 and 271 were issued for Hatch, Units 1 and 2, respectively, to adopt Technical Specification Task Force (TSTF) Traveler TSTF-591, Revise Risk Informed Completion Time (RICT) Program, and revise the TSs Section 5.5.16, Risk Informed Completion Time Program, to reference Regulatory Guide 1.200, Revision 3, instead of Revision 2, January 28, 2025 (ML24365A040).

The NRC staffs review of the Vogtle, Units 1 and 2, PRAs were based on the NRC staffs previous determination that the PRA models were found acceptable to support issuance of the above amendments regarding Implementation of Topical Report Nuclear Energy Institute NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specification (RMTS) Guidelines, Revision 0 dated August 8, 2017 (ML15127A669); and use of 10 CFR 50.69 dated December 17, 2014 (ML14237A034).

The NRC staffs review of the Farley, Units 1 and 2, PRAs were based on the staffs previous determination that the PRA models were found acceptable to support issuance of amendments regarding implementation of NEI 06-09, Risk Informed Technical Specifications Initiative 4B Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A dated August 23, 2019 (ML19175A243); and to adopt 10 CRF 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors dated June 30, 2021 (ML21137A247).

The NRC staffs review of the Hatch, Units 1 and 2, PRAs were based on the staffs previous determination that the PRA models were found acceptable to support issuance of amendments regarding Revision to Technical Specification to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b dated December 22, 2022 (ML22297A146); and Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors dated June 26, 2020 (ML20077J704).

The relief request notes for Farley and Vogtle, Units 1 and 2, that all open Facts and Observations (F&Os) were found to be closed by an independent F&O Closure Review team per NEI 05-04/07-12/12-06 Appendix X, Close-Out of Facts and Observations (F&Os). For Hatch, the NRC staff reviewed two open F&Os related to accident sequence and human reliability technical elements and found SNC adequately dispositioned these findings with respect to this application.

In its letter dated June 27, 2024, the licensee stated, in part, that, SNC intends to review and assess the existing SNC PRAs to ensure they are of adequate technical capability to support Code Case N-752.

SNC shall review changes to the plant, operational practices, applicable plant and industry operational experience, and, as appropriate, update the PRA and categorization and treatment processes. SNC shall perform this review in a timely manner but no longer than once every two refueling outages. This approach is consistent with the feedback and adjustment process of 10 CFR 50.69(e).

Although the passive methodology proposed in NL-24-0143 is similar to that used in the RI-ISI program, the licensee confirmed that it will continue to review and assess the existing PRAs to demonstrate adequate technical capability and to maintain a feedback and process adjustment process consistent with that of 10 CFR 50.69(e) to update the PRA, categorization, and treatment processes based on review of changes to the plant, operational practices and applicable plant and industry operational experiences. The NRC finds this approach for PRA technical adequacy, feedback and process adjustment to be acceptable.

3.5.2 Active Function Evaluation For pressure retaining components that have a passive function as well as an active function, the proposed alternative categorization process only applies to the pressure boundary function of these components and no treatment changes will be applied to the active function as a resulting of implementing the proposed alternative. However, the consequence evaluation methodology of the proposed alternative must address not only the postulated failure of the subject pressure boundary component (e.g., loss of a flow path) but also other direct and indirect failures including any effects of the active function. Therefore, while treatment requirements for the active portion of the pressure retaining components are not within the scope of the proposed alternative, the assessment of the impact to the active function is required by the proposed plant-specific methodology.

The proposed categorization methodology is the consequence evaluation portion of Electric Power Research Institute (EPRI) TR-112657 Revision B-A, Revised Risk-Informed Inservice Inspection Procedure (ML013470102), which is the foundational methodology for several risk-informed applications related to SSCs that perform pressure boundary functions. These applications include ASME Code Case 660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement ActivitiesSection XI, Division I, RI-ISI programs, and ANO-R&R-004, Revision 1. Relative risk measures such as Fussell-Vesely (F-V) and Risk Reduction Worth (RRW) are not applied for these applications, in part, because passive components and pressure retaining portion of active components typically have very low failure rates/probabilities; and common cause failure probabilities are also very low and would reach orders of magnitude below the truncation levels of the PRA. As such, using relative importance measures such as F-V and RRW identifies the vast majority of pressure boundary components and pressure retaining functions of active components as low safety significant. The F-V and RRW importance measures are often used for the selection of candidates for improvement and enhanced maintenance, whereas the Conditional Core Damage Probability (CCDP) criteria, applied in Code Case N-752-1, and thus SNC specific requests NL-24-0143, is useful for identifying components that should be prevented from failing using repair/replacement, planned maintenance and other treatment requirements. Code Case N-752-1 notes:

Section 1420, item c: Changes in configuration, design, materials, fabrication, examination, and pressure-testing requirements used in the repair/replacement activity shall be evaluated, as applicable, to ensure the structural integrity and leak tightness of the system are sufficient to support the design bases functional requirements of the system.

These requirements, in addition to those outlined in the relief request as explained in this safety evaluation, provide reasonable confidence that passive components and pressure retaining functions of active components will continue to perform their design-basis functions, and, therefore, would not impact the basis for using F-V.

3.5.4 Risk Tables The proposed alternative references Code Case N-752 which allows for the use of risk tables as identified in Table I-1, I-2, I-3 and I-4 in lieu of CCDP or Conditional Large Early Release Probability (CLERP). As explained in Code Case N-752, differences in consequence rank between the use of risk tables and quantitative indices shall be reviewed, justified, and documented or the higher consequence rank assigned. The risk-informed process that will be used by SNC to determine the risk-informed safety classification (RISC) for Class 2 and 3 moderate energy systems is the following:

Scope identification,

Consequence evaluation,

Consequence categorization,

Classification considerations,

Final classification definitions, and

Reevaluation of risk-informed safety classifications.

This methodology includes systematic consideration of initiating events and operating states that may be outside the scope of SNC PRAs such as refueling operations.

3.5.5 Review of Key Principles The NRC staff evaluated the application to the guidance in Regulatory Guide (RG) 1.174 Key Principles. These key principles are:

Principle 1: The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under 10 CFR 50.12, Specific Exemptions).

Principle 2: The proposed licensing basis change is consistent with a defense-in-depth [DID]

philosophy.

Principle 3: The proposed licensing basis change maintains sufficient safety margins.

Principle 4: When proposed 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.

In the submittal dated June 27, 2024, the licensee stated:

Code Case N-752 exempts LSS items, which have been categorized as LSS in accordance with the code case, from having to comply with the repair/replacement requirements of ASME Section XI. Exempted ASME Code requirements for LSS items are outlined in Section 3, above. In lieu of these requirements, Code Case N-752, Paragraph -1420 requires the Owner to define alternative treatment requirements which confirm with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design basis conditions. These Owner treatment requirements must address or include all the provisions stipulated in Paragraphs -1420(a) through (j) of the code case. This approach to treatment is consistent with RISC-3 treatment requirements specified in 10 CFR 50.69(d)(2).

To comply with the above, SNC will develop and/or revise existing procedures and documents to define treatment requirements for performing repair/replacement activities on LSS items in accordance with the Code Case N-752. Defined treatment requirements address design control, procurement, installation, configuration control, and corrective action. These procedures and documents also include provisions which address/implement the following requirements:

1. Administrative controls for performing these repair/replacement activities.
2. The fracture toughness requirements of the original Construction Code and Owners Requirements shall be met.
3. Changes in configuration, design, materials, fabrication, examination, and pressure testing requirements used in the repair/replacement activity shall be evaluated, as applicable, to ensure the structural integrity and leak tightness of the system are sufficient to support the design bases functional requirements of the system.
4. Items used for repair/replacement activities shall meet the Owners Requirements or revised Owners Requirements as permitted by the licensing basis.
5. Items used for repair/replacement activities shall meet the Construction Code to which the original item was constructed. Alternatively, items used for repair/replacement activities shall meet the technical requirements of a nationally recognized code, standard, or specification applicable to that item as permitted by the licensing basis.
6. The repair methods of nationally recognized post construction codes and standards (e.g., PCC-2, API-653) applicable to the item may be used.
7. Performance of repair/replacement activities, and associated non-destructive examination (NDE), shall be in accordance with the Owners Requirements and, as applicable, the Construction Code, or post construction code or standard, selected for the repair/replacement activity. Alternative examination methods may be used as approved by the Owner. NDE personnel may be qualified in accordance with IWA-2300 in lieu of the Construction Code.
8. Pressure testing of the repair/replacement activity shall be performed in accordance with the requirements of the Construction Code selected for the repair/replacement activity or shall be established by the Owner.
9. Baseline examination (e.g., preservice examination) of the items affected by the repair/replacement activity, if required, shall be performed in accordance with requirements of the applicable program(s) specifying periodic inspection of items.

See paragraph 5.2.E.11, below, for additional details.

10. Implementation of Code Case N-752 does not negate or affect SNC commitments to regulatory and enforcement authorities having jurisdiction at HNP Units 1&2, FNP Units 1&2 and VEGP 1&2.
11. Periodic ISI and inservice testing (IST) of LSS items at HNP Units 1&2, FNP Units 1&2 and VEGP 1&2 will be performed as follows:

ISI of LSS pressure retaining items or their associated supports will be performed in accordance with the sites ISI program implemented in accordance with 10 CFR 50.55a.

IST of pumps and valves that have been classified as LSS will be performed in accordance with the sites IST program implemented in accordance with 10 CFR 50.55a.

IST of snubbers that have been classified as LSS will be performed in accordance with the sites Snubber Testing program implemented in accordance with 10 CFR 50.55a.

Inspections of LSS items performed under other plant programs, such as the Flow Accelerated Corrosion and Microbiologically Induced Corrosion programs, will continue to be performed under those programs for the site.

12. Adverse conditions identified in LSS components will be entered in the SNC corrective action program, which satisfies 10 CFR 50 Appendix B criteria for corrective action. Conditions that would prevent an LSS item from performing its safety related function(s) under design basis conditions will be corrected in a timely manner. For SSCs under 10 CFR 50.36, Technical Specifications, adverse conditions will be addressed within the timeline of the limiting conditions of operability, or the necessary action statements will be performed. For significant conditions adverse to quality, measures will be taken to provide reasonable confidence that the cause of the condition is determined, and corrective action taken to preclude repetition. The SNC corrective action process takes appropriate actions to monitor, investigate, and/or correct undesired conditions with the level of emphasis and effort commensurate with the risk and significance of the issue. Finally, this approach to corrective action of LSS items is consistent with the NRC position on corrective action of RISC-3 SSCs as specified in 10 CFR 50.69(d)(2)(ii).
13. As permitted by Code Case N-752, SNC intends to implement the exemption from IWA-1400(f) and IWA-4000 applicable to utilization of an Authorized Inspection Agency (AIA) and Authorized Nuclear Inservice Inspector (ANII) when performing repair/replacement activities on LSS items. In lieu of ANII inspection services, SNC believes that its proposed treatment requirements, as described herein, provide reasonable confidence that LSS systems and items remain capable of performing their safety-related functions when repair/replacement activities are performed without the inspection services of an ANII. It should also be noted that the exemption of ANII services is not unique to Code Case N-752.

Utilization of ANII inspection services is already exempt by ASME Section XI for certain items and activities such as small items (IWA-4131) and rotation of items for testing or preventative maintenance (IWA-4132). Finally, exemption of AIA/ANII services for this code case application is consistent with the NRCs position on risk-informed programs as specified in 10 CFR 50.69(b)(1)(v).

14. Code Case N-752 paragraph -1420 allows LSS items to be exempt from the requirements of certain ASME Section XI, including subparagraph IWA-1400(n) and article IWA-4000. However, Code Case N-752 does not allow exemption from ASME Section XI subparagraph IWA-1400(n) if compliance with 10 CFR 50 Appendix B or NQA-1 is required at the Owner's facility, as is the case for SNC's nuclear facilities. SNC's Quality Assurance Program requirements, currently applicable to Farley, Hatch, and Vogtle Units 1 & 2, are described in the SNC Quality Assurance Topical Report (QATR). Changes to the QATR are subject to the regulatory change control requirements of 10 CFR 50.54(a)(3). Accordingly, SNC intends to amend QATR Section 1.1 Scope / Applicability in accordance with 10 CFR 50.54(a)(3) to include the following: For SNC nuclear sites having received NRC authorization to use the alternative repair/replacement categorization and treatment requirements of ASME Code Case N-752 in lieu of the corresponding sections of ASME Section XI, as referenced in 10 CFR 50.55a Codes and Standards, treatment of safety-related structures, systems, and components identified as low safety significant (LSS) Class 2 and 3 SSCs in accordance with ASME Code Case N-752 is not required to meet the requirements of this manual. Instead, treatment of these LSS SSCs is performed in accordance with existing QA Program procedures and processes which include supplemental controls to ensure the capability and reliability of the SSCs design basis function. The basis for the SNC QATR change is established in the precedent identified in Section 7.2 of this alternative request and in accordance with 10 CFR 50.54(a)(3)(ii), which establishes that a quality assurance alternative or exception approved by an NRC safety evaluation is not considered a reduction in QA Program commitments provided the bases of the NRC approval are applicable to the licensee's facility. Consistent with the precedent in Section 7.2, under the amended QATR, SNC will define alternative treatment requirements that confirm with reasonable confidence that each Class 2 and 3 LSS SSC will remain capable of performing its safety-related function under design-basis conditions. In doing so, SNC will use current QA Program processes and procedures with additional controls for the treatment of Class 2 and 3 LSS components to reasonably assure continued capability and reliability of the design-basis function(s). This includes confirming, with reasonable confidence, that changes to the configuration, design, material, fabrication, examination, and testing requirements used to support repair/replacement activities on Class 2 and 3 LSS SSCs are performed in accordance with SNC's existing design change process and addressing in SNC's corrective action program any condition that may prevent a LSS SSC from performing its design-basis function. For the procurement of Class 2 and 3 LSS components as non-safety-related for repair/replacement activities in accordance with ASME's Code Case N-752, supplemental procurement requirements will be specified, and additional controls will be implemented as appropriate to provide reasonable assurance that Class 2 and 3 LSS SSCs will remain capable of performing their safety-related function under design basis conditions. Such controls include conducting receipt inspections using qualified inspection personnel consistent with SNC's procurement requirements and prohibiting suppliers of Class 2 and 3 SSCs and subparts from making design changes or changes to the procurement order without prior SNC approval. Using these existing QA Program processes and alternative treatment requirements, SNC believes that the implementation of ASME Code Case N-752 will provide reasonable assurance that each Class 2 and 3 LSS SSC remains capable of performing its design-basis function, and the SNC QATR will continue to provide an acceptable level of quality and safety.
15. As permitted by Code Case N-752, SNC intends to implement the exemptions from IWA-1400(j) and IWA-4000 applicable to repair/replacement programs and plans. In lieu of these ASME Section XI administrative controls, SNC will establish Owner defined administrative controls as required by paragraph -1420(a) of Code Case N-752. SNC will utilize its existing work management processes for planning and documenting the performance of repair/replacement activities and supplement those process requirements as necessary to comply with Code Case N-752. These controls will confirm, with reasonable confidence, that repair/replacement activities on LSS items are performed in accordance with work instructions that have been appropriately, planned, reviewed, and implemented. It should also be noted that the exemption of Repair/Replacement Plans as required by IWA-1400(j) and IWA-4150 is not unique to Code Case N-752. Repair/Replacement Plans are already exempt by ASME Section XI for certain items and activities such as small items (IWA-4131) and rotation of items for testing or preventative maintenance (IWA-4132). Finally, the exemption of ASME Section XI programs and plans and the alternative use of Owner-defined administrative requirements on LSS items is consistent with the NRCs position on risk-informed programs as specified in 10 CFR 50.69(b)(1)(v).
16. As permitted by Code Case N-752, SNC intends to implement the exemption on IWA-4000 applicable to repair/replacement activities. Article IWA-4000 of the ASME Section XI Code specifies administrative, technical, and programmatic requirements for performing repair/replacement activities on pressure retaining items and their supports. As specified in IWA-4110(b), repair/replacement activities include welding, brazing, defect removal, metal removal by thermal means, rerating, and removing, adding, and modifying items or systems. These requirements are applicable to procurement, design, fabrication, installation, examination, and pressure testing of items within the scope of this Division. In lieu of these IWA-4000 requirements, SNC will perform repair/replacement activities on LSS items in accordance with an Owner defined program that complies with paragraph -1420 of ASME Code Case N-752. The SNC program will utilize existing SNC processes such as those applicable to procurement, design, re-rating, fabrication, installation, modifications, welding, defect removal, metal removal by thermal processes and supplement those process requirements as necessary to comply with Code Case N-752. SNC believes this program will confirm, with reasonable confidence, that LSS items remain capable of performing their safety-related functions under design basis conditions. Finally, the exemption of IWA-4000 requirements and the alternative use of Owner-defined treatment requirements for LSS items is consistent with the NRCs position on risk-informed programs as specified in 10 CFR 50.69(b)(1)(v) and (d)(2).
17. As permitted by Code Case N-752, SNC intends to implement the documentation exemptions on IWA-6211(d), IWA-6211(e), and IWA-6350. These ASME Section XI paragraphs address preparation and retention of Enclosure to NL-24-0143 Proposed Alternative GEN-ISI-ALT-2024-01 Page 17 of 20 various ASME Section XI records such as Form NIS-2, IWA-4160 verification of acceptability evaluations, IWA-4311 evaluations, Repair/Replacement Plans, and reconciliation documentation. In lieu of these ASME Section XI forms and evaluations, the following repair/replacement activity records shall be retained in accordance with SNCs Owner-defined program for performing repair/replacement activities on LSS items.

Repair/replacement activity documentation.

Evaluations of LSS items that do not comply with requirements of the applicable Construction Code, standard, specification, and/or design specification. See also paragraph 5.2.E.12 above.

Evaluations and documentation of design and configuration changes including material changes.

Based on the above, the NRC staff finds that the licensees adherence to the above elements covered in Code Case N-752-1 for repair/replacement activities provides reasonable confidence that each LSS item will remain capable of performing its safety-related function. The repair/replacement program quality elements will ensure that the LSS items remain capable of performing their design safety function.

While the NRC staff finds that a clearly defined code or standard is preferable for the predictability and clarity of the alternate treatment to be implemented, the NRC staff concludes that the proposed alternative permits acceptable flexibility in treatment alternatives, specifically for Class 2 and 3 LSS components, through a methodology described above and Farley, Units 1 and 2, Hatch, Units 1 and 2, and Vogtle, Units 1 and 2, plant-specific evaluation. Because the proposed alternative treatment is limited to LSS components, with defined treatment requirements (e.g., design control, corrective action, etc.) described in the enclosure to the licensees submittal, the NRC staff finds that the codes and standards, as described, provide an acceptable level of quality and safety.

Key Principle 2:

In the submittal dated June 27, 2024, the licensee stated that its request to use Code Case N-752-1 with no exceptions or deviations, including all definitions. The categorization process described in Code Case N-752-1 includes the consideration of DID. According to Appendix I of Code Case N-752-1, the categorization process demonstrates DID philosophy is maintained if the following requirements in Code Case N-752-1:

Reasonable balance is preserved among prevention of core damage, prevention of containment failure or bypass, and mitigation of an offsite release.

There is no over-reliance on programmatic activities and operator actions to compensate for weaknesses in the plant design.

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.

Potential for common cause failures is taken into account in the risk analysis categorization.

Independence of fission-product barriers is not degraded.

In the submittal dated June 27, 2024, SNC stated, in part:

The ASME Code Case N-752 risk-informed categorization evaluation is performed by an Owner-defined team that includes members with expertise in PRA, plant operations, system design, and safety or accident analysis. The risk-informed categorization process is based on the conditional consequence of failure, given that a postulated failure has occurred. A consequence category for each piping segment or component is determined via a failure modes and effects analysis (FMEA) and impact group assessment. The FMEA considers pressure boundary failure size, isolability of the break, indirect effects, initiating events, system impact or recovery, and system redundancy. The results of the FMEA for each system, or portion thereof, are partitioned into core damage impact groups based on postulated piping failures that (1) cause an initiating event, (2) disable a system/train/loop without causing an initiating event, or (3) cause an initiating event and disable a system/train/loop.

SNC stated that categorization and treatment requirements of Code Case N-752 applicable to repair/replacement activities are consistent with NRC requirements specified in 10 CFR 50.69.

Since the Code Case N-752 categorization process is identical to the 10 CFR 50.69 methodology for pressure boundary components, it captures identical system impacts and results in the same conclusion for passive/pressure boundary components, thereby, the same defense-in-depth philosophy is inherent to Code Case N-752.

The NRC staff finds that the proposed change is consistent with defense-in-depth philosophy.

Key Principle 3:

In the submittal dated June 27, 2024, SNC stated that its request to use Code Case N-752 without exceptions. According to Appendix I of Code Case N-752-1, the categorization process shall verify sufficient margins in engineering analysis and supporting data and margin shall incorporated when determining performance characteristics for Class 2 and 3 SSCs identified as LSS. According to the Code Case N-752-1, sufficient margins are maintained by ensuring that safety analysis acceptance criteria in the plant licensing basis are met, or proposed revisions account for analysis and data uncertainty. If sufficient margins cannot be maintained, the categorization process described in Code Case N-752-1 requires that Class 2 or 3 SSC be identified as high safety significant (HSS), which will continue to meet the requirements of Section XI.

In risk-informed decisionmaking, the guidance in RG 1.174 describes the expectation that licensing basis changes maintain sufficient safety margins. In the 10 CFR 50.69 context, this is ensured through the regulation in 10 CFR 50.69(c)(1)(iv) which requires, in part, reasonable confidence that sufficient safety margins are maintained for SSCs categorized as RISC-3.

The engineering evaluation that will be conducted by the SNC under Code Case N-572 for SSC categorization will assess the design function(s) and risk significance of the SSC to assure that sufficient safety margins are maintained.

The proposed alternative requires the verification of sufficient margin for Class 2 or 3 SSC prior to applying the alternative requirements for LSS. If sufficient margin cannot be verified, then the requirements of Section XI still apply. Therefore, the NRC staff concludes that the proposed change maintains sufficient safety margin and provide an acceptable level of quality and safety.

Key Principle 4:

The passive categorization process is driven by the consequence of failure in that the process conservatively assumes that a failure occurs with a probability of 1.0. As such, some postulated passive failures will be categorized as HSS while, from a pure risk perspective, they may be low safety significant if actual failure frequencies were considered. In addition to modelling incorporated in the PRA which includes direct effects, the methodology addresses indirect effects of the failure including pipe whip, jet impingement, flooding, debris generation and harsh environment. The NRC staff finds that by modelling components as 100 percent failed instead of applying actual failure rates of passive components which are normally 1E-08 or lower; and by including all impacts of the break including both direct and indirect effects into the analysis, the methodology provides a conservative risk assessment of the component.

The NRC staff notes that the proposed changes in treatments are not expected to result in significant changes to existing low failure frequencies and there is reasonable confidence that the affected SSCs would retain the capability and reliability of the design basis function.

Therefore, the NRC staff concludes that the proposed change would result in at most small changes to core damage frequency or risk in accordance with the Commissions Policy Goal statement.

Key Principle 5:

In its letter dated June 27, 2024, SNC described how the impact of the proposed changes would be monitored using performance management strategies.

The licensee stated:

SNC shall review changes to the plant, operational practices, applicable plant, and industry operational experience, and, as appropriate, update the PRA and categorization and treatment processes. SNC shall perform this review in a timely manner but no longer than once every two refueling outages. This approach is consistent with the feedback and adjustment process of 10 CFR 50.69(e).

SNC also stated:

Baseline examination (e.g., preservice examination) of the items affected by the repair/replacement activity, if required, shall be performed in accordance with requirements of the applicable program(s) specifying periodic inspection of items.

The licensee further stated:

Conditions that would prevent an LSS item from performing its safety related function(s) under design basis conditions will be corrected in a timely manner.

For SSCs under 10 CFR 50.36, Technical Specifications, adverse conditions will be addressed within the timeline of the limiting conditions of operability, or the necessary action statements will be performed. For significant conditions adverse to quality, measures will be taken to provide reasonable confidence that the cause of the condition is determined, and corrective action taken to preclude repetition. The SNC corrective action process takes appropriate actions to monitor, investigate, and/or correct undesired conditions with the level of emphasis and effort commensurate with the risk and significance of the issue.

Finally, this approach to corrective action of LSS items is consistent with the NRC position on corrective action of RISC-3 SSCs as specified in 10 CFR 50.69(d)(2)(ii).

Based on the above, the NRC staff concludes that the proposed changes provide reasonable confidence that LSS items would be monitored appropriately using performance management strategies.

3.5.6 Risk Conclusion Based on the above, the NRC staff finds, with reasonable assurance, that the SNC plant-specific PRAs reflect the as-built, as-operated plants to support the safety significance categorization of NL-24-0143, and that the feedback and process adjustments will provide reasonable confidence that the PRA will be maintained in a manner to support the categorization and treatment for the repair/replacement of Class 2 and 3 items. In addition, the NRC staff finds the application to be consistent with RG 1.174 Key Principles.

3.5.7 Quality Assurance Code Case N-752 states LSS items would be exempt from ASME BPV Code,Section XI, IWA-1400(n), which requires the licensee to document repair and replacement activities via a quality assurance (QA) program in accordance with 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, or ASME NQA-1, if Appendix B is not applicable at the nuclear power plant. However, 10 CFR Part 50, Appendix B continues to be required at the subject plants and, therefore, NRC staff review of this alternative is based on compliance with 10 CFR Part 50, Appendix B. In footnote (1), Code Case N-752-1 states, If compliance with 10 CFR 50, Appendix B or NQA-1 is required at the Owners facility, IWA-1400(o) is not exempt (NRC staff notes that the reference of IWA-1400(o) vs. IWA-1400(n) is due to different edition and addenda of the ASME Code, but that the content is the same). For clarity, while the term exempt is used in the cited footnote, the proposed alternative does not exempt the LSS components from 10 CFR Part 50, Appendix B requirements, as any exemption from an NRC regulatory requirement in 10 CFR Part 50 would need to be requested and considered under 10 CFR 50.12 or other more specific provisions, as appropriate. However, the proposed alternative allows for altering the treatment of those LSS components under the provisions of 10 CFR Part 50, Appendix B.

In its submittal dated June 27, 2024, SNC stated:

SNCs Quality Assurance Program requirements, currently applicable to Farley, Hatch, and Vogtle Units 1 & 2, are described in the SNC Quality Assurance Topical Report (QATR). Changes to the QATR are subject to the regulatory change control requirements of 10 CFR 50.54(a)(3).

To support the QA treatment aspects of its request, SNC references an NRC staff safety evaluation (ML21132A279) dated May 19, 2021, approving a proposed change to the Quality Assurance Program Manual (QAPM) at ANO under 10 CFR 50.54(a)(4) with specific QA requirements under 10 CFR Part 50, Appendix B (although less rigorous than previously applied at ANO under 10 CFR Part 50, Appendix B) for safety-related Class 2 and Class 3 components categorized as LSS when implementing Code Case N-752 at ANO. The NRC approval of the changes to the QAPM at ANO are based on the specific QA requirements for safety-related LSS Class 2 and Class 3 components when implementing Code Case N-752 documented in the Entergys submittals for ANO dated October 26, 2020 (ML20300A324), April 5, 2021 (ML21095A244), and April 30, 2021 (ML21120A326).

As part of the review of the QA treatment aspects of the SNC request, the NRC staff confirmed that the QA changes proposed by SNC are consistent with the QA changes approved by the NRC staff for ANO as documented in the safety evaluation dated May 19, 2021 (ML21132A279), therefore, it is not considered a reduction in commitment in accordance with 10 CFR 50.54(a)(3)(ii). The NRC staff concluded that there is reasonable assurance that the licensees QA treatment for safety-related Class 2 and Class 3 components categorized as LSS when implementing Code Case N-752 at the specified SNC Fleet nuclear power plants continue to meet the requirements of 10 CFR Part 50, Appendix B.

In Section 5.1, Overview of Code Case N-752, of the alternative request, SNC stated that the categorization and treatment requirements of Code Case N-752 are consistent with those in 10 CFR 50.69. In the request for additional information (RAI) dated October 16, 2024 (ML24290A155), the NRC staff requested that the licensee describe the provisions of the SNC Fleet categorization and treatment processes that are consistent with 10 CFR 50.69. In its response to RAI-1 dated November 15, 2024 (ML24320A129), SNC stated, this alternative request is not proposing to comply with or seeking alternatives from the categorization requirements of 50.69. In its RAI response, the licensee further stated that, the Owner must specify or establish process requirements applicable to design control, procurement, installation, configuration control, corrective action, etc. to ensure that LSS items remain capable of performing their safety related functions under design basis conditions. Additionally, SNC stated that, ISI and IST [inservice testing] on Class 2 and Class 3 LSS items must continue to comply with the SNC ISI and IST programs implemented in accordance with 10 CFR 50.55a, Codes and standards.

In the RAI dated October 16, 2024, the NRC staff requested the licensee to clarify that QA and treatment activities for LSS safety-related SSCs remain within the scope of 10 CFR Part 50, Appendix B, when implementing Code Case N-752, with risk-informed relaxation of specific activities allowed by the NRC when authorizing N-752 requests, with the design, licensing basis, and regulatory enforcement remaining subject to 10 CFR Part 50, Appendix B, regardless of the change to the SNC QATR [Quality Assurance Topical Report]. In its response to RAI-2 dated November 15, 2024, SNC stated, in part, that, the SNC QATR will allow repair/replacement activities on items categorized as LSS to be in accordance with Code Case N-752 provided treatment of these LSS SSCs is performed in accordance with existing QA Program procedures and processes which include supplemental controls to ensure capability and reliability of the SSC design basis function. These supplemental controls include conducting receipt inspections using qualified inspection personnel consistent with SNC's procurement requirements and prohibiting suppliers of Class 2 and 3 SSCs and subparts from making design changes or changes to the procurement order without prior SNC approval. SNC believes that implementation of this QATR change will provide reasonable assurance that Class 2 and 3 LSS SSCs will remain capable of performing their safety related functions under design basis conditions.

SNC agrees that the treatment provisions of Code Case N-752 on Class 2 and 3 LSS items are required to comply with 10 CFR 50 Appendix B. However, SNC is not proposing an alternative to 10 CFR 50 Appendix B. Rather, SNC is proposing to implement a change to its QATR that would allow use of Code Case N-752 for plants that are authorized to use the Code Case.

In the RAI dated October 16, 2024, the NRC staff requested that the licensee clarify the statements in the SNC Fleet alternative request regarding: (1) the applicability of 10 CFR Part 50, Appendix B, during the use of Code Case N-752, (2) the submittals reference to the use of Code Case N-752 without exception although 10 CFR Part 50, Appendix B, will continue to apply, (3) meeting reasonable confidence vs. reasonable assurance to confirm that Appendix B will continue to be met, and (4) updates to the QATR regarding Class 2 and Class 3 3 LSS items in that these items are not exempt from 10 CFR Part 50, Appendix B. In its response to RAI-3, the licensee reiterated part of its response to RAI-2 and stated, in part, that, SNC is not proposing an exemption from 10 CFR 50 Appendix B. Rather, the proposed alternative would allow for altering the treatment of LSS components under the provisions of 10 CFR 50 Appendix B. SNC also stated that, Although Code Case N-752 takes exception to IWA-1400(o) (IWA-1400(n) of the ASME Section XI Code of record for applicable SNC sites),

footnote 1 in Code Case N-752 states that if compliance with 10 CFR 50 Appendix B or NQA-1, is required at the Owners facility, IWA-1400(o) is not exempt. In addition, the licensee stated, in part, that, SNCs application of the Appendix B controls as outlined in Section 5.2 of the alternative request provides reasonable assurance that LSS items remain capable of performing its safety function under design basis conditions.

Based on the above, the NRC staff finds that the QA and treatment activities for LSS safety-related Class 2 and Class 3 SSCs remain within the scope of 10 CFR Part 50, Appendix B, and provide reasonable assurance that SNCs proposed plant-specific implementation of Code Case N-752, with risk-informed relaxation of specific activities documented in the alternative request, as supplemented, is acceptable. Implementation of this authorization remains subject to 10 CFR Part 50, Appendix B, and verification through the NRC Reactor Oversight Process.

3.6 NRC Staff Conclusion

Based on information provided, the NRC staff finds that: (1) the proposed risk categorization methodology will satisfactorily classify the affected Class 2 and 3 components as HSS or LSS, (2) the alternate treatment requirements in the proposed alternative will provide reasonable assurance that each LSS item remains capable of performing its safety related function, (3) the current RI-ISI program will continue, (4) the licensees corrective action program will continue to provide actions to correct conditions that could prevent an LSS item from performing its safety function, (5) the feedback and process adjustment will allow timely update of the elements of this program, (6) the licensees PRA has sufficient technical quality to support this request, and (7) the repair/replacement program quality elements will provide reasonable assurance that the LSS items remain capable of performing their design safety function. Therefore, the NRC staff finds that the proposed alternative will provide an acceptable level of quality and safety.

4.0 CONCLUSION

Based on the above, the NRC staff determined that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1).

Therefore, the NRC staff authorizes the proposed alternative for the remainder of the fifth ISI interval at Farley, Units 1 and 2, the remainder of the fifth ISI interval at Hatch, Units 1 and 2, and the remainder of the fourth ISI interval at Vogtle, Units 1 and 2.

All other ASME Code,Section XI, requirements for which an alternative was not specifically requested and authorized in this alternative remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

This authorization does not change any obligations regarding the NRC-approved 10 CFR 50.69 programs for risk-informed categorization and treatment for SSCs at Farley, Units 1 and 2, Hatch, Units 1and 2, and Vogtle, Units 1 and 2.

Principal Contributors: Jigar Patel, NRR Thomas Scarbrough, NRR Date: March 27, 2025

ML24198A073 (Package)

ML24198A062 (Letter/SE)

OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-11/LA NRR/DRA/APLA/BC NRR/DRO/IQVB NAME JLamb KZeleznock RPascarelli KKavanagh DATE 01/07/2025 02/05/2025 12/02/2024 02/04/2025 OFFICE NRR/DNRL/NPHP/BC NRR/DNRL/NPHP NRR/DNRL/NPHP NRR/DEX/EMIB NAME MMitchell* (non-concur) JCollins* (non-concur)

JHoncharik*

(non-concur)

TScarbrough* (non-concur)

DATE 02/06/2025 02/06/2025 02/06/2025 02/21/2025 OFFICE NRR/DEX/EMIB/BC NRR/DORL/DD NAME SBailey ARivera-Varona DATE 02/06/2025 03/27/2025

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