ML26013A184
| ML26013A184 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  (NPF-042) |
| Issue date: | 02/06/2026 |
| From: | Michael Mahoney Plant Licensing Branch IV |
| To: | Reasoner C Wolf Creek |
| Lee S, 301-415-3168 | |
| References | |
| EPID L-2025-LLR-0051 | |
| Download: ML26013A184 (0) | |
Text
February 6, 2026 Mr. Cleveland Reasoner Chief Executive Officer and Chief Nuclear Officer Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839
SUBJECT:
WOLF CREEK GENERATING STATION, UNIT 1 - AUTHORIZATION OF PROPOSED ALTERNATIVE I5R01 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 (EPID L-2025-LLR-0051)
Dear Mr. Reasoner:
By letter dated May 8, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25128A304), as supplemented by letter dated September 24, 2025 (ML25267A244), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for authorization to use proposed alternative I5R01 to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section XI, requirements at Wolf Creek Generating Station, Unit 1 (Wolf Creek).
Specifically, the licensee requested 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, in lieu of certain requirements in ASME BPV Code Section XI, articles IWA-1000, IWA-4000, and IWA-6000, as incorporated by reference in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, Codes and standards.
Pursuant to 10 CFR 50.55a(z)(1), the licensee 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 submittal as a plant-specific request for Wolf Creek.
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 of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the proposed alternative for the fifth inservice inspection interval, which begins on September 3, 2025, and ends on September 2, 2037.
C. Reasoner 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.
In its request, the licensee mentioned in several places that risk-informed categorization process described in ASME Code Case N-752 is consistent with the risk-informed categorization process described in 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. 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 Samson Lee at 301-415-3168 or via email at Samson.Lee@nrc.gov.
Sincerely, Michael Mahoney, Acting Branch Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482
Enclosure:
Safety Evaluation cc: Listserv MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.02.06 10:54:07 -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 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482
1.0 INTRODUCTION
By letter dated May 8, 2025 (Agencywide Documents Access and Management System Accession No. ML25128A304), as supplemented by letter dated September 24, 2025 ML25267A244), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) requested an alternative to certain requirements of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (BPV Code) 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, in lieu of certain requirements in ASME BPV Code,Section XI, articles IWA 1000, IWA-4000, and IWA-6000, as incorporated by reference in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, Codes and standards, at Wolf Creek Generating Station, Unit 1 (Wolf Creek).
Specifically, pursuant to 10 CFR 50.55a(z)(1), the licensee 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 submittal as a plant-specific request for Wolf Creek.
2.0 REGULATORY EVALUATION
2.1 Regulations The following requirement is applicable to this request:
10 CFR 50.55a(z)(1), Acceptable level of quality and safety 2.2 Regulatory Guidance The NRC staff used the following guidance in the evaluation of this request:
NRC Information Notice 2025-01, Lessons Learned When Implementing ASME Code Case N-752, dated February 10, 2025 (ML24323A057)
3.0 TECHNICAL EVALUATION
3.1 Licensees Proposed Alternative 3.1.1 Licensees Applicable ASME BPV Code Edition and Addenda The applicable ASME BPV Code editions and addenda for the specific inservice inspection (ISI) intervals are specified in the table below.
Plant ISI Interval ASME BPV Code Section XI, Code of Record Edition/Addenda Interval Start Interval End Wolf Creek 5th 2019 Edition 9/3/2025 9/2/2037 3.1.2 ASME BPV Code Components Affected Section 1.0, ASME Code Components Affected, of the attachment to the licensees alternative request letter dated May 8, 2025, lists the affected Code components, which are ASME BPV Code Class 2 and 3 components and their supports except Class CC and MC items, greater than Nominal Pipe Size (NPS) 4 piping within the break exclusion region, and greater than NPS 4 feedwater system piping from the steam generator, including the steam generator, to the outer containment isolation valve.
3.1.3 Applicable ASME BPV Code Requirements ASME BPV Code,Section XI, subsection IWA, provides the requirements for repair/replacement activities. Section 3.0, Applicable Code Requirements, of the attachment to the licensees alternative request letter dated May 8, 2025, provides additional details of the Code requirements related to group classification, arrangement with an Authorized Inspection Agency, written programs and plans, documentation of a Quality Assurance Program, specific requirements for pressure-retaining items or their supports, Owner reporting responsibilities, and Owner retained records.
3.1.4 Proposed Alternative Section 5.0, Proposed Alternative and Basis for Use, of the attachment to the licensees alternative request letter dated May 8, 2025, states:
Pursuant to 10 CFR 50.55a(z)(1), WCNOC proposes to implement ASME Code Case N-752 as an alternative to the ASME Code requirements specified in Section 3. 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. 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 ASME Code Case N-752 will provide an acceptable level of quality and safety.
3.2
NRC Staff Evaluation
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 ISI requirements in the ASME BPV Code,Section XI, as incorporated by reference in 10 CFR 50.55a. The NRC staff verified that the precedents cited by the licensee are applicable.
The NRC staff also reviewed deviations from the licensee-identified precedents.
3.2.1 Applicability of Precedents The licensee provided a list of precedents in section 7.0 of the attachment to the licensees alternative request letter dated May 8, 2025. The precedents are Arkansas Nuclear One, Units 1 and 2 (ANO), alternative EN-20-RR-001, approved May 19, 2021 (ML21118B039); and Oconee Nuclear Station, Units 1, 2, and 3 (Oconee), alternative RA-22-0174, approved December 13, 2023 (ML23262A967). The NRC staff did not re-review these precedents, but verified that they are applicable to this licensee request.
Section 5.2.B, Categorization Process, of the attachment to the licensees alternative request letter dated May 8, 2025, indicates that the licensees categorization process is technically equivalent to the process approved by the NRC staff on April 22, 2009, for ANO2-R&R-004, Revision 1 (ML090930246). Alternative ANO2-R&R-004 was the basis for ASME Code Case N-752. 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 ASME Code Case N-752. This process requires technical acceptability of the probabilistic risk assessment (PRA) used to support the ASME Code Case N-752 required evaluations as described in section 3.2.3 of this safety evaluation.
Section 5.2.B of the attachment to the licensees alternative request letter dated May 8, 2025, indicates that the licensees 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. Section 5.2.D, Feedback and Process Adjustment, indicates that the licensee will continue to review and assess the existing PRAs to demonstrate adequate technical capability and to maintain a feedback and 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. Although Wolf Creek does not have an approved 10 CFR 50.69 program, the NRC staff finds this approach for categorization, feedback, and process adjustment to be acceptable.
Therefore, the NRC staff verified that the approved ANO2-R&R 004, Revision 1, categorization methodology is applicable to Wolf Creek.
Section 5.2.E, Treatment Requirements for LSS [Low Safety Significant] Items, of the attachment to the licensees alternative request letter dated May 8, 2025, states, in part:
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 WCNOCs existing corrective action program.
Section 5.2.E of the attachment to the licensees alternative request letter dated May 8, 2025, indicates that the licensees alternative treatment requirements include specific provisions. The NRC staff found that the ANO alternative EN-20-RR-001, and Oconee alternative RA-22-0174, precedents have similar provisions for alternative treatment. Therefore, the Wolf Creek provisions for alternative treatment are acceptable. In addition, section 5.2.E of the attachment to the licensees alternative request letter dated May 8, 2025, states, in part:
WCNOC will develop alternative elements under the provisions of Appendix B describing treatment of these LSS SSCs [systems, structures, and components]
to ensure continued capability and reliability of the design basis function. The procedures governing these treatment activities are classified as safety-related and are therefore, under the jurisdiction of 10 CFR 50, Appendix B.
The NRC staff finds that the quality assurance (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, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, and provide reasonable assurance that the licensees proposed plant-specific implementation of ASME Code Case N-752, with risk-informed relaxation of specific activities documented in the alternative request, as supplemented, is acceptable. The NRC staff also finds that it is consistent with NRC Information Notice 2025-01. Implementation of this authorization remains subject to 10 CFR Part 50, Appendix B, and verification through the NRC Reactor Oversight Process.
The NRC staff notes that the ANO licensee followed 10 CFR 50.54, Conditions of licenses, paragraph (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 ASME Code Case N-752. The Wolf Creek licensee proposes to follow 10 CFR 50.54(a)(3) to apply 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 Wolf Creek licensee must satisfy the QA treatment allowed by the NRC staff for the ANO licensee request under 10 CFR 50.54(a)(4) for components within the scope of ASME Code Case N-752 to provide an acceptable level of quality and safety under 10 CFR 50.55a(z)(1).
3.2.2 Deviations from Precedents The NRC staff did not identify any significant deviation from the licensee cited precedents.
3.2.3 PRA Technical Acceptability The NRC staff verified the technical acceptability of the Wolf Creek PRA model for use in the risk-informed inservice inspection (RI-ISI) program, consistent with the requirements in ASME Code Case N-752. The NRC staffs review of the Wolf Creek PRA supporting the RI-ISI program was partly based on the NRC staffs previous determinations that the PRA model was acceptable to support the issuance of Wolf Creek Amendment No. 227 regarding adoption of Technical Specification Task Force (TSTF) Traveler TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk Informed TSTF] Initiative 5b, dated April 8, 2021 (ML21053A117). As indicated in section 5.2 of the attachment to the licensees alternative request letter dated May 8, 2025, the licensees peer review assessed the internal events, including internal flooding, PRA model against ASME/ANS RA-Sa-2009, Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addenda to RA-S-2008, February 2009, and Regulatory Guide (RG) 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML090410014). The licensee indicated that there are no remaining open finding-level facts and observations (F&Os). The final closure of finding-level F&Os was documented in November 2023 and was performed in accordance with NRC-approved processes as documented in section 3.3 of the WCNOC 10 CFR 50.69 license amendment request (ML25030A384). Based on the above, the NRC staff finds the Wolf Creek PRA acceptable to use for the purpose of this proposed alternative to allow use of ASME Code Case N-752.
3.3 NRC Staff Conclusion
The NRC staff finds the licensee cited precedents that are applicable to Wolf Creek regarding the use of ASME Code Case N-752. The proposed use of N-752 in this application is consistent with previous alternatives to ASME BPV Code,Section XI, requirements that NRC staff has previously concluded that did meet the five key principles of risk-informed decision-making in RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ML17317A256). The NRC staff also did not identify any significant deviation from the cited precedents. The NRC staff finds that the licensees PRA is of sufficient quality and is acceptable for this application. The feedback and process adjustment will allow timely update of the elements of this program. 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 fifth ISI interval at Wolf Creek.
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.
Principal Contributors: Drew Childs, NRR Thomas Scarbrough, NRR Todd Hilsmeier, NRR Yiu Law, NRR Date: February 6, 2026
- by eConcurrence **by email OFFICE NRR/DORL/LPL4/PM*
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NRR/DRA/APLA/BC(A)** NRR/DRO/IQVB/BC**
NAME SLee PBlechman MGonzalez KKavanagh DATE 1/14/2026 1/28/2026 12/16/2025 12/23/2025 OFFICE NRR/DNRL/NPHP/BC** NRR/DEX/EMIB/BC**
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NAME MMitchell SBailey MMahoney DATE 12/19/2025 12/22/2025 2/06/2026