ML25128A304
| ML25128A304 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/08/2025 |
| From: | Boyce M Wolf Creek |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| Download: ML25128A304 (1) | |
Text
P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 Michael T. Boyce Vice President Engineering May 8, 2025 000873 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Docket No. 50-482: Proposed Alternative to Use American Society of Mechanical Engineers (ASME) Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1 Commissioners and Staff:
Pursuant to 10 CFR 50.55a(z)(1), Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests Nuclear Regulatory Commission (NRC) approval of a proposed alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components for WCNOCs Inservice Inspection (ISI) Program.
Specifically, WCNOC is requesting to use the alternative requirements of ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 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 in lieu of certain ASME Code Section XI, paragraph IWA-1000, IWA-4000, and IWA-6000 requirements. WCNOC requests approval on the basis that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1). The Attachment provides the basis for the request.
WCNOC requests approval by April 30, 2026, to allow for planning upcoming refueling outages.
This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-8831 x8687, or Dustin Hamman at (620) 364-4204.
Sincerely, Michael T. Boyce MTB/jkt
000873 Page 2 of 2
Attachment:
Request for Alternative in Accordance with 10 CFR 50.55a(z)(1) to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1" cc:
A. N. Agrawal (NRC), w/a S. S. Lee (NRC), w/a J. D. Monninger, (NRC), w/a Senior Resident Inspector (NRC), w/a WCNOC Licensing Correspondence ET 25-000873, w/a
Attachment to 000873 Page 1 of 15 Wolf Creek Nuclear Operating Corporation Wolf Creek Generating Station Relief Request Number I5R01 Request for Alternative in Accordance with 10 CFR 50.55a(z)(1) to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1
Attachment to 000873 Page 2 of 15 1.0 ASME CODE COMPONENTS AFFECTED:
This request applies to ASME Class 2 and 3 items or components except the following:
- 1. Piping within the break exclusion region [> Nominal Pipe Size (NPS) 4 (DN 100)] for high energy piping systems1 as defined by the Owner.
- 2. 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.
- 3. This request does not apply to Class CC and MC items.
1 NUREG 0800 Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, provides a method for defining the scope of piping.
2.0 APPLICABLE CODE EDITION AND ADDENDA:
Wolf Creek Generating Station (WCGS) is currently in its fourth Inservice Inspection (ISI) interval which started on September 3, 2015, and is scheduled to end on September 2, 2025. However, the proposed alternative is being requested for the fifth ISI interval, which is scheduled to begin on September 3, 2025, and is scheduled to end on September 2, 2037. The edition of the ASME Code Section XI which will be in use for the fifth ISI interval will be the 2019 Edition (Reference 8.1).
Per Regulatory Issue Summary (RIS) 2004-12 (Reference 8.1), Letter ET 20-0011 (Reference 8.12) and approval (Reference 8.3), the NRC staff concluded that the use of subparagraph IWA-4540(b) of the 2017 Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, is acceptable for WCGS.
3.0 APPLICABLE CODE REQUIREMENTS:
3.1. 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) requires Owners to possess or obtain an arrangement with an Authorized Inspection Agency (AIA).
IWA-1400(k) requires Owners to perform repair/replacement activities in accordance with written programs and plans.
IWA-1400(o) requires Owners to maintain documentation of a Quality Assurance Program in accordance with 10 CFR 50 or ASME NQA-1, 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 Report for Repair/Replacement Activity.
Attachment to 000873 Page 3 of 15 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.
4.0 REASON FOR REQUEST:
Wolf Creek Nuclear Operating Corporation (WCNOC) will be performing repair/replacement activities at WCGS during the fifth ISI interval in accordance with a deterministic Repair/Replacement Program based on the 2019 Edition of ASME Section XI. Section XI Repair/Replacement requirements apply to procurement, design, fabrication, installation, examination, and pressure testing of items within the scope of ASME Section XI.
Repair/replacement activities include welding, brazing, defect removal, metal removal using thermal processes, rerating, and removing, adding, or modifying pressure-retaining items or supports. Repair/replacement activities are performed in accordance with WCNOC 10 CFR 50, Appendix B Quality Assurance (QA) Program and the ASME Section XI Code. In applying a deterministic approach to repair/replacement activities, a safety class (e.g.,
ASME Class 2 or 3) is assigned to every component within a system based on system function; the same treatment requirements are then applied to every component within the system without considering the risk associated with the probability that a specific item or component may or may not be functional at a time when needed.
Alternatively, a probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, Probabilistic Risk Assessment (PRA) addresses credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner. In 2004, the NRC adopted a new Section 50.69 of 10 CFR relating to risk-informed categorization and treatment of structures, systems, and components (SSCs) for nuclear power plants (Reference 8.4).
This new section permits power reactor licensees to implement an alternative regulatory framework with respect to special treatment (treatment beyond normal industrial practices) of low safety significant (LSS) SSCs. In May 2006, the NRC staff issued Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, For Trial Use, Revision 1 (Reference 8.5). RG 1.201 endorses a categorization method, with conditions, for categorizing active SSCs described in Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline.
WCNOC is not requesting NRC approval to implement 10 CFR 50.69 in this relief request.
Instead, WCNOC is proposing to implement the risk-informed categorization and treatment requirements of ASME Code Case N-752 when performing repair/replacement activities on Class 2 and 3 pressure-retaining items or their associated supports. Code Case N-752, which was approved by the ASME in July 2019, employs a comprehensive categorization process requiring input from both a PRA model and deterministic insights. This approach will enable evaluation, categorization, and implementation of alternative treatments for resolution of emergent issues in segments of piping having low safety significance. Use of Code Case N-752 will also allow WCNOC to identify and more clearly focus engineering,
Attachment to 000873 Page 4 of 15 maintenance, and operations resources on critical components with high safety-significance, thus, enabling WCNOC to make more informed decisions and increase the safety of the plant.
5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:
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 Code Case N-752 will provide an acceptable level of quality and safety.
5.1. Overview of Code Case N-752 Code Case N-752 provides for risk-informed categorization and treatment requirements for performing repair/replacement activities on Class 2 and 3 pressure-retaining items or their associated supports. Code Case N-752 is not applicable to the following:
Class CC and MC items.
Piping within the break exclusion region [> 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 PWRs from the SG, including the SG, to the outer containment isolation valve.
Code Case N-752 categorization methodology relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, the risk-informed process categorizes components solely based on consequence, which measures the safety significance of the component given that it ruptures (component failure is assumed with a probability of 1.0). This approach is conservative compared to including the rupture frequency in the categorization as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to changes in treatment. It additionally applies deterministic considerations (e.g., defense in depth, safety margins) in determining safety significance. Additional detail is provided Section 5.2.
The risk-informed process categorizes components as either high safety significant (HSS) or LSS. HSS components must continue to meet ASME Section XI rules for repair/replacement activities. LSS components are exempt from ASME Section XI repair/replacement requirements and can be repaired/replaced in accordance with treatment requirements established by the Owner. The treatment requirements must provide reasonable confidence that each LSS item remains capable of performing its safety-related functions under design basis conditions. Component supports, if categorized, are assigned the same safety significance, HSS or LSS, as the highest
Attachment to 000873 Page 5 of 15 passively ranked segment within the bounds of the associated analytical pipe stress model. The categorization and treatment requirements of Code Case N-752 are consistent with those in 10 CFR 50.69.
It should be noted that Code Case N-752 is based on ANO-2 relief request ANO2-R&R-004, Revision 1, dated April 17, 2007 (Reference 8.6), as supplemented by Entergy. The NRC approved relief request ANO2-R&R-004, Revision 1, in a safety evaluation dated April 22, 2009 (Reference 8.7). The ANO-2 relief 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.
5.2. Basis for Use The information below is provided as a basis or justification for WCNOC proposed alternative to implement the risk-informed categorization and treatment requirements of Code Case N-752 on Class 2 and 3 pressure-retaining items or the associated supports as defined in Section 1.
A.
Application to Individual Items Within a System The risk-informed methodology of Code Case N-752 may be applied on a system basis or on individual items within selected systems. Paragraph -1100 of Code Case N-752 states: This Case may be applied on a system basis, including all pressure-retaining items and their associated supports, or on individual items categorized as low-safety-significant (LSS) within the selected systems. While this is the case, the risk-informed methodology is, in actuality, applied to the pressure boundary function of the individual components within the system. The risk-informed methodology contained in Code Case N-752 requires that the components pressure boundary function be assumed to fail with a probability of 1.0, and all impacts caused by the loss of the pressure boundary function be identified. This would include identifying impacts of the pressure boundary failure on the component under evaluation, identifying impacts of the pressure boundary failure of the component on the system in which the component resides, as well as identifying impacts of the pressure boundary failure of the component on any other plant SSC. This includes direct effects (e.g. loss of the flow path) of the component failure and indirect effects of the component failure (e.g. flooding, spray, pipe whip, loss of inventory). This comprehensive assessment of total plant impact caused by a postulated individual component failure is then used to determine the final consequence ranking. As such, the final consequence rank of the individual component would be the same regardless of whether the entire system or only the individual component is subject to the risk-informed methodology.
B.
Categorization Process The categorization process of Code Case N-752 is delineated in Appendix I of the Code Case. This categorization process is technically identical to the process approved by the NRC under Relief Request ANO2-R&R-004, Revision 1 (Reference 8.6), which, in turn, is based on founding principles in EPRI Report TR-112657, Revision B-A, Revised Risk-Informed Inservice Inspection
Attachment to 000873 Page 6 of 15 Evaluation Procedure, and the categorization process of Code Case N-660, but with improvements and lessons learned from trial applications.
The Code Case N-752 risk-informed categorization evaluation is performed by an Owner-defined team that includes experts 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 cause an (1) 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.
Failures are also evaluated for their importance relative to containment performance. In addition, the consequence rank is reviewed and adjusted to reflect the pressure boundary failures impact on plant operation during shutdown and on the mitigation of external events. Credit may be taken for plant features and operator actions to the extent these would not be adversely affected by failure of the piping segment or component under consideration.
Consequence evaluation results are ranked as High, Medium, Low, or None (no change to base case). Piping segments/components ranked as High by the consequence evaluation process are considered HSS and require no further review. Piping segments/components ranked as Medium, Low, or None by the consequence evaluation shall be determined to be HSS or LSS by evaluating the additional categorization considerations or conditions outlined in paragraph I3.4.2(b) of Code Case N-752. If any of these conditions are not met, then HSS shall be assigned. If all conditions are met, then LSS may be assigned. Finally, if LSS is assigned, the categorization process shall verify that there are sufficient margins to account for uncertainty in the engineering analysis and supporting data. If sufficient margin exists, then LSS should be assigned. If sufficient margin does not exist, then HSS shall be assigned.
C.
PRA Technical Adequacy The following demonstrates that the quality and level of detail of the processes used in categorization of SSCs are adequate. The PRA models described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed.
The PRA models credited in the License Amendment Wolf Creek Generating Station, Unit 1 - Issuance of Amendment No. 227 Re: Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Based on TSTF-425 (Reference 8.8) are indicated below. No methodology upgrades have been used relative to internal events or internal flood for this risk informed application which would have required a focused peer review consistent with
Attachment to 000873 Page 7 of 15 Regulatory Guide (RG) 1.200 Revision 2 (Reference 8.16).
Wolf Creek Generating Station Internal Events Probabilistic Risk Assessment Model-of-Record (MOR) 10. This WCNOC Model-of-Record incorporated maintenance only updates to the previous MOR9 (peer reviewed as described below). The updates included the addressing of findings that remained open from the peer review of MOR9. MOR9 was previously accepted by NRC for TSTF 425 (Reference 8.8) with disposistions of the peer review findings that remained open at the time of submittal. A subsequent F&O closure effort resulted in closing all remaining findings with MOR10.
Wolf Creek Generating Station Internal Flooding Probabilistic Risk Assessment - Based on Internal Event PRA MOR9. It is also noted that this WCNOC MOR was previously accepted by NRC for TSTF 425 (Reference 8.8). Updates of the internal flooding model result in the use of interim model 9.3 used for N-752 purposes.
The WCGS Code Case N-752 categorization process for internal events (IE) and internal flooding (IF) hazards use the plant-specific WCNOC PRA model which has been used on previous risk informed applications and was determined to be technically acceptable.
The WCNOC risk management process ensures that the PRA MOR used in this application reflects the as-built and as-operated plant for WCGS. The WCNOC process delineates the responsibilities and guidelines for updating and maintaining the PRA models current with the design and operation of the station and includes criteria for both regularly scheduled and interim PRA model updates.
The PRA hazard models described above have been assessed against RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 8.16).
The current IE (MOR10) and IF (MOR9) have been peer reviewed and have no Findings remaining open against them. The initial peer review of PRA technical adequacy was for the 2019 WCGS Internal Events PRA (Reference 8.19). This peer review was performed during the week of June 17-21, 2019. The assessment evaluated the Internal Events PRA (including internal flooding) against the requirements published in the current version of the ASME/ANS PRA Standard (RA-Sa-2009) (Reference 8.15) and RG 1.200 (Reference 8.16).
This independent peer review was followed up with an initial facts and observation closure effort (Reference 8.17). Final closure of remaining facts and observations was documented in (Reference 8.18).
The above information demonstrates that the PRA is of sufficient quality and level of detail to support the categorization process and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.
Attachment to 000873 Page 8 of 15 D.
Feedback and Process Adjustment The WCNOC process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, equipment performance, errors or limitations identified in the model, and industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. WCNOC 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. WCNOC shall perform this review in a timely manner but no longer than once every two complete refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated and presented to the Integrated Decision-making Panel (IDP).
This approach is consistent with the feedback and adjustment process of 10 CFR 50.69(e).
E.
Treatment Requirements for LSS Items 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 of 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, WCNOC will develop new and/or revise existing procedures and documents to define treatment requirements for performing repair/replacement activities on LSS items in accordance with Code Case N-752. WCNOC defined treatment requirements will address design control, procurement, installation, configuration control, and corrective action. WCNOC procedures and documents will 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
Attachment to 000873 Page 9 of 15 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 WCNOC commitments to regulatory and enforcement authorities having jurisdiction at WCGS.
ISI of LSS pressure-retaining items or their associated supports will be performed as required in accordance with WCGSs 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 WCGSs 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 WCGSs 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 will continue to be performed under those programs for WCGS.
- 12. Conditions that would prevent an LSS item from performing its safety-related function(s) under design basis conditions will be corrected in a
Attachment to 000873 Page 10 of 15 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. Finally, this approach to corrective action of LSS items is consistent with the NRC position on corrective action of Risk-Informed Safety Class (RISC)-3 SSCs as specified in 10 CFR 50.69(d)(2)(ii).
- 13. As permitted by Code Case N-752, WCNOC intends to implement the exemption on IWA-1400(g) and IWA-4000 applicable to utilization of an AIA and Authorized Nuclear Inservice Inspector (ANII) when performing repair/replacement activities on LSS items. In lieu of ANII inspection services, WCNOC 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 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. As permitted by Code Case N-752, WCNOC intends to implement the QA Program exemption applicable to IWA-1400(o) and IWA-4000 when performing repair/replacement activities on LSS items. That said, this code case exemption only applies if compliance with 10 CFR 50, Appendix B, or NQA-1 is not required by the NRC at the Owners facility. To address this issue, WCNOC will update the Wolf Creek Quality Program Manual (WCQPM) for safety-related Class 2 and 3 SSCs identified as LSS in accordance with ASME Code Case N-752 to not be required to meet the requirements of the WCQPM. WCNOC will develop alternative elements under the provisions of Appendix B describing treatment of these LSS SSCs 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. In accordance with 10 CFR 50.54(a)(3)(ii), WCNOC is not requesting prior NRC approval of the change to the WCQPM because the NRC has previously determined that this type of change does not constitute a reduction in commitment in accordance with 10 CFR 50.54(a)(3)(ii), as evidenced by the following.
The NRC previously approved a change to the Entergy Quality Assurance Program Manual (Reference 8.9) in conjunction with a request for Arkansas Nuclear One to adopt ASME Code Case N-752 (References 8.10 and 8.11)
The NRCs approval of the subject relief request for Duke Energy (Reference 8.14), stated, in part that The NRC staff confirmed that the changes to the QAPD proposed by Duke Energy are consistent
Attachment to 000873 Page 11 of 15 with the changes approved by the NRC staff to Entergys QAPM as documented in the SE dated May 19, 2021, therefore, it is not considered a reduction in commitment in accordance with 10 CFR 50.54(a)(3)(ii).
- 15. As permitted by Code Case N-752, WCNOC intends to implement the exemptions on IWA-1400(k) and IWA-4000 applicable to repair/replacement programs and plans. In lieu of these ASME Section XI administrative
- controls, WCNOC will establish Owner-defined administrative controls as required by paragraph -1420(a) of Code Case N-752. WCNOC 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 ensure 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(k) 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, WCNOC 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, WCNOC will perform repair/replacement activities on LSS items in accordance with an Owner-defined program that complies with paragraph -1420 of Code Case N-752. The WCNOC program will utilize existing WCNOC 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. WCNOC believes this program will ensure, 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).
Attachment to 000873 Page 12 of 15
- 17. As permitted by Code Case N-752, WCNOC 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 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 WCNOCs 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.
Evaluations and documentation of design and configuration changes including material changes.
In addition to the above, WCNOC will also revise applicable WCGS licensing basis documents (e.g., Updated Safety Analysis Report), as appropriate, to identify systems, subsystems, or individual items that have been categorized as LSS and address alternative treatment requirements. Changes to licensing basis documents will be performed in accordance with 10 CFR 50.59.
F.
Conclusion Code Case N-752 specifies requirements for performing risk-informed categorization and treatment for performing repair/replacement activities on Class 2 and 3 pressure-retaining items or associated supports. The Code Case N-752 categorization process provides a comprehensive methodology for determining the safety significance of items - HSS or LSS. This categorization process is technically identical to that approved by the NRC under relief request ANO2-R&R-004, Revision 1 (Reference 8.6). Repair/replacement activities performed on items determined to be HSS or uncategorized items must continue to comply with the ASME Section XI Code.
Repair/replacement activities performed on LSS items may comply with alternative treatment requirements that are defined by the Owner but must comply with all provisions of paragraph -1420 of Code Case N-752. WCNOCs proposed treatment requirements, as described herein, meet these criteria, and provide reasonable confidence that LSS systems and items remain capable of performing their safety-related functions under design basis conditions. Finally, 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.
6.0 DURATION OF PROPOSED ALTERNATIVE The proposed alternative is being requested for use during the fifth inservice inspection interval for WCGS which begins on September 3, 2025, and ends on September 2, 2037.
The fifth interval will be implemented as a 12-year interval per Code Case N-921 as conditioned by NRC RG 1.147, Revision 21.
Attachment to 000873 Page 13 of 15 7.0 PRECEDENT 7.1. Entergy Operations, Inc., Arkansas Nuclear One Units 1 and 2 Request for Relief No.
EN-20-RR-001, submitted May 27, 2020 (ADAMS Accession No. ML20148M343),
approved May 19, 2021 (ADAMS Accession No. ML21118B039).
7.2. Duke Energy Oconee Nuclear Station Unit 1, 2, and 3, Request for Relief Number RA-22-0174, submitted July 27, 2022 (ADAMS Accession No. ML22208A031), and approved December 13, 2023 (ADAMS Accession No. ML23262A967)
Attachment to 000873 Page 14 of 15
8.0 REFERENCES
8.1. American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, 2019 Edition.
8.2. NRC Regulatory Issue Summary, 2004-12, Clarification on use of Later Editions and Addenda to the ASME OM Code and Section XI, July 28, 2004 (ADAMS Accession No. ML042090436).
8.3. NRC letter to Mr. C. Reasoner, WCNOC, "Wolf Creek Generating Station, Unit 1, Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI (EPID L-2020-LLR-0134),
October 30, 2020 (ADAMS Accession No. ML20302A08).
8.4. 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, And Components for Nuclear Power Reactors, USNRC, 69 FR 68047, November 22, 2004.
8.5. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, And Components in Nuclear Power Plants According to Their Safety Significance, May 2006.
8.6. Entergy Letter to NRC dated April 17, 2007, Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate Energy Systems, (ADAMS Accession No. ML071150108) as supplemented by letters August 6, 2007 (ADAMS Accession No. ML072220160), February 20, 2008 (ADAMS Accession No. ML080520186), and January 12, 2009 (ADAMS Accession No. ML090120620).
8.7. Safety Evaluation (SE) by the Office of Nuclear Reactor Regulation Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems, April 22, 2009 (ADAMS Accession No. ML090930246).
8.8. NRC Letter to Mr. C. Reasoner, Wolf Creek Generating Station, Unit 1 - Issuance of Amendment No. 227 Re: Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Based on TSTF-425 (EPIDL-2020-LLA-0091), April 8,2021 (ADAMS Accession No. ML21053A117).
8.9. NRC Letter to Entergy, Arkansas Nuclear One, Units 1 and 2 - Request for Approval of Change to the Entergy Quality Assurance Program Manual, May 19, 2021 (ADAMS Accession No. ML21132A279).
8.10. Entergy letter to NRC, Relief Request Number EN-20-RR-001 - 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, May 27, 2020 (ADAMS Accession No. ML20148M343).
8.11. NRC Letter to Entergy, Arkansas Nuclear One, Units 1 and 2 - Approval of Request for Alternative from Certain Requirements of The American Society of Mechanical
Attachment to 000873 Page 15 of 15 Engineers Boiler and Pressure Vessel Code, May 19, 2021 (ADAMS Accession No. ML21118B039).
8.12. WCNOC Letter ET 20-0011 to NRC, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code,Section XI, October 1, 2020 (ADAMS Accession No. ML20280A32).
8.13. Duke Energy Letter to NRC, Request for Alternative in Accordance with 10 CFR 50.55a(z)(1) to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 System Section XI, Divisions 1, July 27, 2022 (ADAMS Accession No. ML22208A031).
8.14. NRC Letter to Duke Energy, Oconee Nuclear Station, Units 1, 2, and 3 - RE:
Authorization of Alternative to use RR-22-0174, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1 (EPID: L-2022-LLR-0060), December 13, 2023 (ADAMS Accession No. ML23262A967).
8.15. 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.
8.16. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009.
8.17. PWROG-19038-P, Revision 0, Independent Assessment of Facts & Observations Closure of the Wolf Creek Probabilistic Risk Assessment, March 2020 8.18. PWROG-23024-P, Revision 0, Independent Assessment of Facts & Observations Closure of the Wolf Creek Probabilistic Risk Assessments, November 2023.
8.19. WCNOCPES029-REPT-001, Wolf Creek Internal Events Probabilistic Risk Assessment Peer Review, Revision 0, September 9, 2019.