ML26013A042
| ML26013A042 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Davis Besse, Perry, 07201043, 07200069 |
| Issue date: | 02/02/2026 |
| From: | Ilka Berrios Plant Licensing Branch III |
| To: | Hamilton D Energy Harbor Nuclear Corp |
| Kuntz, RF | |
| References | |
| EPID L-2025-LLR-0043 | |
| Download: ML26013A042 (0) | |
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February 2, 2026 Mr. Ken Peters Executive Vice President and Chief Nuclear Officer Vistra Operations Company LLC 6322 N FM 56 P.O. Box 1002 Glen Rose, TX 76043
SUBJECT:
BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2; COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2; DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AND PERRY NUCLEAR POWER PLANT, UNIT NO. 1-AUTHORIZATION OF PROPOSED ALTERNATIVE 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-2025-LLR-0043)
Dear Mr. Peters:
By letter dated April 1, 2025 (Agencywide Documents Access and Management System Accession No. ML25091A263), as supplemented by letter dated October 6, 2025 (ML25279A012),Vistra Operations Company LLC (Vistra OpCo, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for authorization to use the proposed alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section XI, requirements at Beaver Valley Power Station (Beaver Valley), Unit Nos. 1 and 2, Comanche Peak Nuclear Power Plant (Commanche Peak),
Unit Nos. 1 and 2, Davis-Besse Nuclear Power Plant, Unit No. 1 (Davis-Besse), and Perry Nuclear Power Plant, Unit No. 1 (Perry). 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 BPV Code Section XI, 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) 50.55a(z)(1), Vistra OpCo 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 Beaver Valley, Unit Nos. 1 and 2, Commanche Peak, Unit Nos. 1 and 2, Davis-Besse, Unit No.1 and Perry, Unit No. 1.
The NRC staff has reviewed the proposed alternative, and, as set forth in the enclosed safety evaluation, the NRC staff has 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 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 current renewed operating licenses which is January 29, 2036, for Beaver Valley, Unit No. 1, May 27,
K. Peters 2047, for Beaver Valley Unit No. 2, February 8, 2050, for Commanche Peak, Unit No. 1, February 2, 2053, for Commanche Peak Unit 2, April 22, 2037, for Davis-Besse, Unit No., and November 7, 2046, for Perry, Unit No.1.
All other ASME BPV 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.
If you have any questions, please contact the Vistra OpCo Fleet Senior Project Manager Robert Kuntz at 301-415-3733 or Robert.Kuntz@nrc.gov.
Sincerely, Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-334, 50-412, 50-445, 50-446, 50-346, and 50-440
Enclosure:
Safety Evaluation cc: Listserv ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2026.02.02 11:37:57 -05'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED ALTERNATIVE TO USE ASME CODE CASE N-752, RISK-INFORMED CATEGORIZATION AND TREATMENT FOR REPAIR/REPLACEMENT ACTIVITIES IN CLASS 2 AND 3 SYSTEMS, SECTION XI, DIVISION 1 BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 PERRY NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NOS. 50-334, 50-412, 50-346, AND 50-440
1.0 INTRODUCTION
By letter dated April 1, 2025 (Agencywide Documents Access and Management System Accession No. ML25091A263), as supplemented by letter dated October 6, 2025 (ML25279A012), Vistra Operations Company LLC (Vistra OpCo, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for authorization to use proposed alternative VISTRA-ISI-ALT-2024-01 to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section XI, requirements at Beaver Valley Power Station (Beaver Valley), Unit Nos. 1 and 2, Comanche Peak Nuclear Power Plant (Commanche Peak), Unit Nos. 1 and 2, Davis-Besse Nuclear Power Plant, Unit No. 1 (Davis-Besse), and Perry Nuclear Power Plant, Unit No. 1 (Perry). 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-6212, IWA-6220, and IWA-6350.
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), Vistra OpCo 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 Beaver Valley, Unit Nos. 1 and 2, Commanche Peak, Units 1 and 2, Davis-Besse, and Perry.
2.0 REGULATORY EVALUATION
2.1 Regulations The following requirements are applicable to this request:
10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants 10 CFR 50.55a(z)(1), Acceptable level of quality and safety 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 1.174, 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 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009 (ML090410014)
NRC Information Notice 2025-01, Lessons Learned When Implementing ASME Code Case N-752, February 10, 2025 (ML24323A057)
3.0 TECHNICAL EVALUATION
3.1 The Licensees Proposed Alternative The applicable ASME BPV Code editions and addenda for the applicable inservice inspection (ISI) intervals are specified in the table below for each plant.
Plant ISI Interval ASME BPV Code,Section XI, Code of Record Edition/Addenda Interval Start Interval End Beaver Valley, Unit No. 1 5th 2013 08/29/2018 08/28/2028 Beaver Valley, Unit No. 2 4th 2013 08/29/2018 08/28/2028 Commanche Peak, Unit No. 1 4th 2019 08/13/2020 08/12/2030 Commanche Peak, Unit No. 2 4th 2019 08/03/2023 08/02/2033 Davis-Besse, Unit No.1 5th 2017 06/08/2023 09/20/2032 Perry, Unit No. 1 4th 2013 05/18/2019 05/17/2029 3.2 ASME BPV Code Components Affected In Section 1 of Attachment TXX-24069 to its letter dated April 1, 2025, Vistra OpCo stated:
This request applies to American Society of Mechanical Engineers (ASME)
Class 2 and 3 items or components except the following:
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.
This request does not apply to Class (CC) Concrete Containment and Class (MC) Metal Containment items.
3.3 Applicable ASME BPV Code Requirements In Section 3 of Attachment TXX-24069 to its letter dated April 1, 2025, Vistra OpCo 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(g)2 requires Owners to possess or obtain an arrangement with an Authorized Inspection Agency (AIA).
IWA-1400(k)2 requires Owners to perform repair/replacement activities in accordance with written programs and plans.
IWA-1400(o)2 requires Owners to maintain documentation of a Quality Assurance Program in accordance with 10 CFR 50 or ASME NQA-1, Part I.
IWA-4000 specifies requirements for performing ASME Section XI repair/replacement activities on pressure-retaining items or their support.
IWA-6211(d)2, IWA-6211(e)2, and IWA-62202 specify Owner reporting responsibilities such as preparing Form NIS-2, Owner's Report for Repair/Replacement Activity.
IWA-6211(f)2 and IWA-6212 specify Repair/Replacement Organization documentation and certification requirements.
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.
2 Code Case N-752 is based on the 2017 Edition of ASME Section XI while Vistra OpCo's Code of Record for CPNPP and CPNPP-2 [Commanche Peak, Units 1 and 2] is the 2019 Edition, DBNPS [Davis-Besse] is the 2017 Edition, and PNPP [Perry], BVPS-1, and BVPS-2 [Beaver Valley, Unit Nos. 1 and 2] is the 2013 Edition of ASME Section XI, Division 1. Paragraph references in Section 3, above, are applicable to the 2013, 2017, and 2019 Editions of ASME Section XI with one exception: IWA-6220 in the 2017 Edition does not exist in or apply to the 2013 Edition.
3.4 Proposed Alternative In Section 5 of Attachment TXX-24069 to its letter dated April 1, 2025, Vistra OpCo stated:
Pursuant to 10 CFR 50.55a(z)(1), Vistra OpCo proposes to implement ASME Code Case N-752 with no exceptions or deviations. ASME Code Case N-752 would be used as an alternative to the ASME Code requirements specified in Section 3 of this request. 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 delineated in Section 1 of this request. This requested implementation includes the categorization of passive SCCs (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 Vistra OpCos 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. In evaluating the Vistra OpCos proposed alternative, the NRC staff considered the past precedent of previous NRC plant-specific approvals related to risk-informed treatment of structures, systems and components (SSCs) for nuclear power plants. Specifically, by letter dated April 22, 2009 (ML090930246), the NRC staff authorized the licensee for Arkansas Nuclear One (ANO), Units 1 and 2, 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. 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 pressure-retaining SSCs.
3.5.1 Probabilistic Risk Assessment Acceptability The proposed plant-specific approach for Vistra OpCo uses 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 (or acceptability) 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 licensing actions:
NRC letter dated April 9, 2004, approved the use of RI-ISI Program for Beaver Valley (ML040780805).
Amendment Nos. 292 and 179 were issued for Beaver Valley, to modify the technical specifications (TSs) to adopt Technical Specification Task Force (TSTF)-425, Revision 3, Relocate Surveillance Frequencies to Licensee-Control -Risk Informed Technical Specification Task Force Initiative 5b (ML14322A461).
Amendment Nos. 301 and 190 were issued for Beaver Valley to adopt National Fire Protection Association (NFPA) Standard 805 (NFPA 805) (ML17291A081 and ML18065A403).
Amendment Nos. 183 and 183 were issued for Commanche Peak, to modify the TSs to adopt the risk-informed completion time by adopting TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 by letter dated August 22, 2022 (ML22192A007).
Amendment Nos 187 and 187, were issued for Commanche Peak, on June 10, 2024, to adopt 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (ML24120A363).
Amendment No. 301 was issued for Davis-Besse, on November 3, 2020, to relocate specific surveillance frequencies to a licensee-controlled program based on TSTF-425 (ML20280A827).
Amendment No. 298 was issued for Davis-Besse, on June 21, 2019, to adopt NFPA 805 (ML19100A306).
Amendment No. 171 was issued for Perry, on February 23, 2016, to adopt TSTF-425, Relocate Surveillance Frequencies to Licensee Control (ML15307A349).
The NRC staffs review of the Beaver Valley PRAs acceptability was based on the NRC staffs previous determination that the PRA models were found acceptable to support issuance of the above relief request to use RI-ISI and amendment regarding implementation of TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee-Control -Risk Informed Technical Specification Task Force Initiative 5b. In its submittal, the licensee states that an independent F&O closure assessment was performed in accordance with Nuclear Energy Institute 05-04/07-12/12-06 Appendix X, Close-Out of Facts and Observations (F&Os) (Appendix X)
(ML17086A431). This process for F&O closure was accepted by the NRC in the letter dated May 3, 2017 (ML17079A427). For Beaver Valley, the NRC staff reviewed the two open F&O findings for the internal events and the internal flooding PRAs and the associated dispositions.
The disposition of the first F&O finding, HR-PR-001, stated that the F&O primarily dealt with documentation for the human reliability analysis and was found to be acceptable with regard to its NFPA 805 application. The staff finds the resolution of this F&O is also applicable to this application and is therefore acceptable. The licensee dispositioned the second F&O finding, IFQU-A5-01, explaining that it is associated with HR-PR-001 and related documentation issues.
The NRC notes that this F&O is outside the scope of this application because the associated supporting requirement is related to human induced flooding. Therefore, the NRC staff finds that Vistra OpCo adequately dispositioned the Beaver Valley PRA F&O findings with respect to this application.
The NRC staffs review of the Commanche Peak PRAs acceptability was based on the staffs previous determination that the PRA models were found acceptable to support issuance of amendments regarding implementation of TSTF-505 and to adopt 10 CFR 50.69. The relief request notes for Commanche Peak that the relevant F&Os findings were closed using the Appendix X process. The NRC confirmed, based on the previous submittals, that all relevant F&Os have been adequately dispositioned for this application.
The NRC staffs review of the Davis-Besse PRAs acceptability was based on the staffs previous determination that the PRA models were found acceptable to adopt TSTF-425 and NFPA 805. In its submittal, the licensee states that F&Os were closed using the Appendix X process. For Davis-Besse, the NRC staff reviewed the one remaining open F&O finding for the internal events and internal flooding PRA models and the associated disposition. The NRC staff finds that Vistra OpCo adequately dispositioned this finding since the supporting requirements associated with the open F&O was shown to meet Capability Category I, which meets the minimum requirements for this application.
The NRC staffs review of the Perry PRAs acceptability was based on the staffs previous determination that the PRA models were found acceptable to adopt TSTF-425. For Perry, the submittal states that all finding level F&Os for the internal events and internal flooding PRAs were closed using the Appendix X process.
Based on the above, the NRC finds that the Vistra OpCo plants meet the requirements for PRA acceptability with regard to this application and that closed peer review F&O findings were closed using an NRC-accepted process.
In its letter dated April 1, 2025, the licensee stated, in part, that, Vistra OpCo 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. Vistra OpCo 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 the attachment to L-24-211 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.
3.5.1.1 Review of Key Principles The NRC staff evaluated the proposed alternative against the five key principles of risk-informed decision-making in Regulatory Guide (RG) 1.174. 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 the 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.
Key Principle 1:
In order to address Key Principle 1, the NRC staff reviewed Section 5.2(E) of the licensees submittal. The NRC staff finds that the licensee confirms that the elements covered in Code Case N-752 for repair/replacement activities provides reasonable confidence that each low-safety-significant (LSS) item will remain capable of performing its safety-related function.
The NRC staff notes that ASME Code Case N-752, Paragraph -1420 requires the Owner to define alternative treatment requirements that will ensure with reasonable confidence 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 the ASME Code Case N-752 methodology.
Because the proposed alternative treatment is limited to LSS components, with Owner defined treatment requirements (e.g., design control, corrective action, etc.) the NRC staff finds that the alternative codes and standards, as described, provide an acceptable level of quality and safety.
Key Principle 2:
In the submittal dated April 1, 2025, the licensee stated that its request to use Code Case N-752 with no exceptions or deviations. The categorization process described in Code Case N-752 includes the consideration of DID. According to Appendix I of Code Case N-752, the categorization process demonstrates DID philosophy is maintained if the following requirements in Code Case N-752:
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 April 1, 2025, the licensee 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.
The proposed alternative does not alter any SSCs and will have no effect on layers of defense, or system redundancy. Additionally, the proposed alternative requires that DID philosophy be maintained. Therefore, the NRC staff concludes that the proposed change is consistent with the DID philosophy.
Key Principle 3:
In the submittal dated April 1, 2025, the licensee stated that it requests to use Code Case N-752 without exceptions. According to Appendix I of Code Case N-752, 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, 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 requires that Class 2 or 3 SSC be identified as high safety significant (HSS), which will continue to meet the requirements of Section XI.
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 provides 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 not be risk significant if actual failure frequencies were considered. In addition to modeling 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 modeling components as failed instead of applying actual failure rates of passive components 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 April 1, 2025, Vistra OpCo described how the impact of the proposed changes would be monitored using performance management strategies.
The licensee stated:
Vistra OpCo 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. Vistra OpCo shall perform this review in a timely manner but no longer than once every two refueling outages.
Vistra OpCo 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 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. Corrective action of adverse conditions associated with LSS items will be identified and addressed in accordance with the existing Vistra OpCo corrective action program that is applicable for each site.
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.
3.5.1.2 Risk Conclusion Based on the above, the NRC staff finds, with reasonable assurance, that the Vistra OpCo plant-specific PRAs reflect the as-built, as-operated plants to support the safety significance categorization of VISTRA-ISI-ALT-2024-01, 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.
3.5.2 Quality Assurance In Alternative Request VISTRA-ISI-ALT-2024-01, the licensee proposes to implement ASME Code Case N-752 with no exceptions or deviations pursuant to 10 CFR 50.55a(z)(1). 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 within the scope of Code Case N-752 and categorized as LSS. The licensee requests the use of this proposed alternative on the basis that the requirements in Code Case N-752 will provide an acceptable level of quality and safety. The NRC staff evaluated the licensees request to determine if the proposed alternative meets the requirements of 10 CFR 50.55a(z)(1) for an acceptable level of quality and safety in lieu of the applicable ISI requirements in the ASME BPV Code,Section XI, as incorporated by reference in 10 CFR 50.55a.
In its submittal, the licensee provides a list of precedents where the NRC staff has authorized the use of ASME Code Case N-752 at other nuclear power plants. Those precedents include Alternative Request ANO2-R&R-004, Revision 1, submitted by Entergy for Arkansas Nuclear One, Units 1 and 2 (ANO) in a letter dated April 17, 2007 (ML071150108), as supplemented by letters dated August 6, 2007 (ML072220160), February 20, 2008 (ML080520186), and January 12, 2009 (ML0901206). The Vistra OpCo licensee states that Code Case N-752 is based on ANO2-R&R-004. The NRC authorized ANO2-R&R-004, Revision 1, in a safety evaluation dated April 22, 2009 (ML090930246). The Vistra OpCo licensee states that the ANO-2 request was developed to serve as an industry pilot for implementing a risk-informed repair/replacement process that included a risk-informed categorization process and treatment requirements.
ASME 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 and replacement requirements of ASME BPV Code,Section XI. 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. The licensee states that it will develop new or revise existing procedures and documents to define treatment requirements for performing repair and replacement activities on LSS items in accordance with Code Case N-752. The licensee states that the defined treatment requirements will address design control, procurement, installation, configuration control, and corrective action. In its submittal, the licensee lists the provisions that will be addressed in the plant-specific procedures and documents for implementation of Code Case N-752.
Among the listed provisions to be addressed when implementing N-752, the licensee states that Code Case N-752 does not exempt Owners from ASME BPV Code,Section XI subparagraph IWA-1400(o) if compliance with 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. or the ASME NQA-1 Standard is required at the Owners facility, which is the case for the Vistra OpCo nuclear facilities. The NRC staff notes that NRC Information Notice 2025-01, Lessons Learned When Implementing ASME Code Case N-752, dated February 10, 2025 (ML24323A057), informs licensees of observed inconsistencies between the language in licensee programs during the implementation of Code Case N-752 and the risk-informed methods the NRC approved to be acceptable to satisfy the requirements of 10 CFR Part 50, Appendix B. These inconsistencies could lead to a misinterpretation that the requirements of 10 CFR Part 50, Appendix B, no longer apply to safety-related Class 2 and Class 3 SSCs that are categorized as LSS when implementing Code Case N-752. IN 2025-01 provides important clarification regarding Appendix B to 10 CFR Part 50 requirements, and is appropriate and acceptable for the licensee to include the information in its alternative treatment. Therefore, the NRC staff finds that 10 CFR Part 50, Appendix B, continues to be applicable when implementing N-752 at the Vistra OpCo nuclear power plants.
The licensee states that it will amend the applicable quality assurance (QA) program requirements for nuclear power plants applicable to this request in accordance with 10 CFR 50.54(a)(3) to address the performance of risk-informed repair/replacement activities in accordance with Code Case N-752. The licensee states that the basis for the QA Program change is established in the ANO precedent identified in its 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 ANO precedent under the amended QA Programs, Vistra OpCo 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. Vistra OpCo 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 functions. 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 Class 3 LSS SSCs are performed in accordance with the licensees design change process and addressing in the licensees corrective action program any condition that may prevent an LSS component from performing its design-basis function.
With respect to the Vistra OpCo licensees reference to the ANO precedent, the NRC staff notes that the ANO licensee followed 10 CFR 50.54(a)(4) to establish supplemental processes and procedures that the NRC staff accepted as satisfying the requirements of 10 CFR Part 50, Appendix B, given the low safety significance of the components within the scope of Code Case N-752. The Vistra OpCo licensee proposes to follow 10 CFR 50.54(a)(3) to apply risk-informed QA treatment of components within the scope of ASME Code Case N-752. The NRC regulations in 10 CFR 50.54(a)(3) allow licensees to adjust their QA activities as accepted by the NRC in response to prior licensee requests under 10 CFR 50.54(a)(4). Therefore, the Vistra OpCo licensee must satisfy the QA treatment allowed by the NRC staff for the ANO licensee request under 10 CFR 50.54(a)(4) to provide an acceptable level of quality and safety under 10 CFR 50.55a(z)(1) for components within the scope of ASME Code Case N-752 when implementing the proposed alternative at BVPS-1, BVPS-2, CPNPP-1, CPNPP-2, DBNPS, and PNPP.
Based on its review, the NRC staff finds that the proposed alternative will provide an acceptable level of quality and safety with respect to QA and the treatment of LSS components within the scope of ASME Code Case N-752 at the Vistra fleet nuclear power plants.
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 licensee has confirmed that the alternate treatment requirements in the proposed alternative will provide reasonable confidence 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 confidence 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 current license which is January 29, 2036, for Beaver Valley, Unit No. 1, May 27, 2047, for Beaver Valley Unit No. 2, February 8, 2050, for Commanche Peak, Unit No. 1, February 2, 2053, for Commanche Peak Unit No. 2, April 22, 2037, for Davis-Besse, Unit No. 1, and November 7, 2046, for Perry, Unit No. 1.
All other ASME BPV 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.
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 obligations regarding the NRC-approved 10 CFR 50.69 programs for risk-informed categorization and treatment for SSCs.
Principal Contributors: Alexander Schwab, NRR Thomas Scarbrough, NRR John Honcharik, NRR Jay Collins, NRR Yamir Diaz-Castillo, NRR Date: February 2, 2026
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