RA-24-0231, Proposed Alternative to Use American Society of Mechanical Engineers Code Case N-752, Risk-Informed Categorization & Treatment for Repair

Proposed Alternative to Use American Society of Mechanical Engineers Code Case N-752, Risk-Informed Categorization & Treatment for Repair

ML24320A015
Person / Time
Site:	Mcguire, Catawba, Harris, Brunswick, Robinson, McGuire
Issue date:	11/15/2024
From:	Ellis K Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-24-0231
Download: ML24320A015 (1)

Text

Kevin M. Ellis General Manager Nuclear Regulatory Affairs, Policy &

Emergency Preparedness Duke Energy 13225 Hagers Ferry Rd., MG011E Huntersville, NC 28078 843-951-1329 Kevin.Ellis@duke-energy.com 10 CFR 50.55a RA-24-0231 November 15, 2024 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 / Renewed License Nos. NPF-35 and NPF-52 McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 / Renewed License Nos. NPF-9 and NPF-17 Brunswick Steam Electric Plant, Units 1 and 2 Docket Nos. 50-325 and 50-324 / Renewed License Nos. DPR-71 and DPR-62 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63 H.B. Robinson Steam Electric Plant, Unit 2 Docket No. 50-261 / Renewed License No. DPR-23

Subject:

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 Pursuant to 10 CFR 50.55a(z)(1), Duke Energy Carolinas, LLC, and Duke Energy Progress, LLC, collectively referred henceforth as Duke Energy, requests the U.S. Nuclear Regulatory Commissions (NRC) authorization of a proposed alternative to the ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components for Catawba Nuclear Station, Units 1 and 2 (CNS), McGuire Nuclear Station, Units 1 and 2 (MNS), Brunswick Steam Electric Plant, Units 1 and 2 (BNP), Shearon Harris Nuclear Power Plant, Unit 1 (HNP), and H.B. Robinson Steam Electric Plant, Unit 2 (RNP).

Specifically, Duke Energy 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 Class 2 and 3 items in lieu of certain ASME Code Section XI, paragraph IWA-1000, IWA-4000, and IWA-6000 requirements. Duke Energy 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).

U.S. Nuclear Regulatory Commission RA-24-0231 Page 2 The proposed alternative is provided in the Enclosure to this submittal. Duke Energy requests NRC approval of the proposed alternative within one year of acceptance.

No regulatory commitments are contained in this submittal.

If there are any questions or if additional information is needed, please contact Mr. Ryan Treadway, Director - Nuclear Fleet Licensing, at 980-373-5873.

Kevin M. Ellis General Manager - Nuclear Regulatory Affairs, Policy & Emergency Preparedness

Enclosure:

cc:

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" L. Dudes, USNRC Region II - Regional Administrator D. Rivard, USN RC Senior Resident Inspector - CNS C. Safouri, USNRC Senior Resident Inspector-MNS G. Smith, USNRC Senior Resident Inspector - BNP P. Boguszewski, USNRC Senior Resident Inspector-HNP J. Zeiler, USNRC Senior Resident Inspector-RNP N. Jordan, NRR Project Manager - Duke Energy Fleet

Enclosure Duke Energy Carolinas, LLC and Duke Energy Progress, LLC (Duke Energy)

Catawba Nuclear Station, Units 1 and 2 McGuire Nuclear Station, Units 1 and 2 Brunswick Steam Electric Plant, Units 1 and 2 Shearon Harris Nuclear Power Plant, Unit 1 H.B. Robinson Steam Electric Plant, Unit 2 Relief Request Number RA-24-0231 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

Relief Request RA-24-0231 Enclosure Page 2 of 32 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. Class CC and MC items.

Notes:

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

2.0 APPLICABLE CODE EDITION AND ADDENDA:

Table 1 Plant/Unit(s)

ISI Interval ASME Section XI Code Edition/Addenda Interval Start Date Interval End Date1 Brunswick Nuclear Plant (BNP) Units 1 and 2 Fifth 2007 Edition, Through 2008 Addendum 3, 4, 5, 6 05/11/2018 05/10/2028 Catawba Nuclear Station (CNS) Units 1 and 2 Fourth 2007 Edition, Through 2008 Addendum3, 4, 5, 6 08/19/2015 06/28/2026 H.B. Robinson Nuclear Plant (RNP) Unit 2 Sixth 2017 Edition6 02/19/2023 02/18/2033 McGuire Nuclear Station (MNS) Unit 1 Fifth 2019 Edition2 12/01/2021 11/30/2031 McGuire Nuclear Station (MNS) Unit 2 Fifth 2019 Edition 03/01/2024 02/28/2034 Shearon Harris Nuclear Plant (HNP) Unit 1 Fourth 2007 Edition, Through 2008 Addendum3, 4, 5, 6 09/09/2017 09/08/2027 Notes:

1. The Interval End Date is subject to change in accordance with IWA-2430(c)(1) or an approved alternative.

2. ISI Examinations and Pressure Testing performed in the first Period are in accordance with Section XI, 2007 Edition with the 2008 Addenda per Relief Request RA-20-0031 (Reference 8.35).

Relief Request RA-24-0231 Enclosure Page 3 of 32

3. Per RIS 2004-12, Letter RA-20-0262, ADAMS Accession Number ML20260H325 and approval ADAMS Accession Number ML20300A206, the NRC staff concluded that the use of subparagraph IWA-4540(b) of the 2017 Edition of the ASME B&PV Code,Section XI, is acceptable. Therefore, as the applicability for this alternative, the Code of Record for IWA-4540(b) is the 2017 Edition (Reference 8.5).

4. Per RIS 2004-16, Letter RA-20-0263, ADAMS Accession Number ML20260H326 and approval ADAMS Accession Number ML21113A013, the NRC staff concluded that the use of subparagraph IWA-4340 of the 2017 Edition of the ASME B&PV Code,Section XI, is acceptable. Therefore, as the applicability for this alternative, the Code of Record for IWA-4340 is the 2017 Edition (Reference 8.7).

5. Per RIS 2004-12, Letter RA-20-0191, ADAMS Accession Number ML20265A028 and approval NRC Accession Number ML21029A335, the NRC staff concluded that the use of paragraphs IWA-5120, IWA-5213, IWA-5241, IWA-5242, and IWA-5250 of the 2017 Edition of the ASME B&PV Code,Section XI, is acceptable. Therefore, as the applicability for this alternative, the Code of Record for IWA-5120, IWA-5213, IWA-5241, IWA-5242, and IWA-5250 is the 2017 Edition (Reference 8.36).

6. Per RIS 2004-12, Letter RA-23-0001, ADAMS Accession Number ML23033A037 and approval ADAMS Accession Number ML23118A076, the NRC staff concluded that the use of paragraph IWA-6230(b) of the 2019 Edition of the ASME B&PV Code,Section XI, is acceptable. Therefore, as the applicability for this alternative, the Code of Record for IWA-6230(b) is the 2017 Edition (Reference 8.37).

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(f)1 requires Owners to possess or obtain an arrangement with an Authorized Inspection Agency (AIA).

IWA-1400(j)1 requires Owners to perform repair/replacement activities in accordance with written programs and plans.

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

Relief Request RA-24-0231 Enclosure Page 4 of 32 IWA-4000 specifies requirements for performing ASME Section XI repair/replacement activities on pressure-retaining items or their supports.

IWA-6211(d)1 and (e)1, specify Owner reporting responsibilities such as preparing Form NIS-2, Owners Report for Repair/Replacement Activity.

IWA-62121 repeats the requirement for certification by a repair/replacement organization and refers to Appendix T as an example.

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

IWA-6350 specifies that the following ASME Section XI repair/replacement activity records must be retained by the Owner:

evaluations required by IWA-4160 and IWA-4311, Repair/Replacement Programs and Plans, reconciliation documentation, and NIS-2 Forms.

NOTES:

1. Code Case N-752 is based on the 2017 Edition of ASME Section XI while Duke Energys Code of record for BNP Units 1 and 2, CNS Units 1 and 2, and HNP Unit 1 is the 2007 Edition/2008 Addenda, except as noted in Section 2.0 of this request. The code of record for McGuire Nuclear Station is the 2019 Edition, except as noted in Section 2.0 of the request. Below is a cross reference for affected code paragraphs for those sites currently adhering to the 2007 Edition/2008 Addenda:

IWA-1400(g), (k), and (o) in the 2017 and 2019 Editions are IWA-1400(f), (j), and (n) in the 2007 Edition/2008 Addenda.

IWA-6211(d) and (e) in the 2017 and 2019 Editions are IWA-6210(d) and (e) in the 2007 Edition/2008 Addenda.

IWA-6212 in the 2017 Edition does not exist in or apply to the 2007 Edition/2008 Addenda. However, IWA-6212 does apply to the 2019 Edition.

4.0 REASON FOR REQUEST:

Duke Energy currently performs repair/replacement activities at BNP Units 1 and 2, CNS Units 1 and 2, RNP Unit 2, MNS Units 1 and 2, and HNP Unit 1, in accordance with a deterministic Repair/Replacement Program based on the

Relief Request RA-24-0231 Enclosure Page 5 of 32 current ASME Section XI code of record as referenced in Table 1.

Repair/Replacement Program 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 Duke Energys 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 8.8).

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 8.9). 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."

Duke Energy is not requesting NRC approval to implement 10 CFR 50.69 in this relief request. Instead, Duke Energy 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 Duke Energy to identify and more clearly focus engineering, maintenance, and operations resources on critical components with high safety-significance, thus, enabling

Relief Request RA-24-0231 Enclosure Page 6 of 32 Duke Energy 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), Duke Energy 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

Relief Request RA-24-0231 Enclosure Page 7 of 32 its safety-related functions under design basis conditions. Component supports, if categorized, are assigned the same safety significance, HSS or LSS, as the highest 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&R004, Revision 1, dated April 17, 2007 (Reference 8.10), as supplemented by Entergy. The NRC approved relief Request ANO2-R&R-004, Revision 1, in a safety evaluation dated April 22, 2009 (Reference 8.11). 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 Duke Energy's 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.

Relief Request RA-24-0231 Enclosure Page 8 of 32 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 8.10), which, in turn, is based on founding principles in EPRI Report TR-112657, Revision B-A, "Revised Risk-Informed Inservice Inspection 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.

Relief Request RA-24-0231 Enclosure Page 9 of 32 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.

Harris The PRA models credited are the same PRA models credited in the following NRC-approved License Amendments:

Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment No. 184 Regarding Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF Initiative 4B (Reference 8.14)

Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment No. 174 Re: Adopt Title 10 of the Code of Federal Regulations 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors" (Reference 8.13)

Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Regarding Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program (Reference 8.15)

The Harris Code Case N-752 categorization process for the internal events and flooding hazard uses the plant-specific PRA model.

The Duke Energy risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the Harris unit.

The PRA 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 8.31).

The HNP internal events PRA model was subject to a self-assessment and a full-scope peer review conducted in 2002 in accordance with guidance in NEI- 00-02 (Reference 8.28), Industry PRA Peer Review Process. In 2006, a self-assessment was conducted to identify supporting requirements that did not meet Category II of the ASME Standard RA-Sb-2005 (Reference 8.29) and RG 1.200 Revision 1 (Reference 8.30). In 2007, a focused scope industry peer review against two elements was conducted as a follow up to the self-assessment against AMSE Standard RA-Sb-2005 and RG 1.200 Revision 1. In July 2017, a focused scope industry peer review was conducted against one model area that was upgraded.

Relief Request RA-24-0231 Enclosure Page 10 of 32 The HNP Internal Events PRA model was subject to a focused-scope peer review conducted in September 2019. The scope included HLRs IE, AS, SC, SY, QU, and LE conducted to the ASME/ANS RA-Sa-2009 (Reference 8.33) PRA standard with NRC clarifications from RG 1.200 Revision 2. This peer review, combined with the focused scope peer review conducted in 2007, form the review of record for the HNP Internal Events PRA model and cover all HLRs in the ASME/ANS RA-Sa-2009 PRA standard, superseding the 2002 peer review.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 (Reference 8.32) in the Internal Events PRA model.

The HNP internal flood PRA model was subject to a self-assessment and a full-scope (covering all internal flood SRs) peer review conducted in August 2014 against RG 1.200 Revision 2. There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Flood PRA model.

Finding level F&Os were reviewed and closed in March 2017 for the Internal Events and Internal Flood models as a pilot for the process documented in the draft of Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) published at the time of the review. NRC staff observed the pilot closure on-site event held January 31 through February 1, 2017. An assessment has been performed to determine the impact of changes to the guidance between the closure event and the final version endorsed by NRC.

The main deltas identified are related to 1) utility and review teams documented determination and justification if each finding resolution is an upgrade verses maintenance update, and 2) the assessment teams confirmation that for the closed F&Os, the aspects of the underlying SRs in ASME/ANS RA-Sa-2009 that were previously not met, or met at CC-I, are now met or met at CC-II. The utility portion of the upgrade verses maintenance assessment was completed globally and did not identify any resolutions as an upgrade. Additionally, the review team determined none of the resolutions were upgrades and this is documented in the final report. The assessment team confirmed resolution of the findings allowed re-categorization of capability categories to meet or met at CC-II, as applicable.

Finding level F&Os generated from the Internal Events September 2019 focused-scope peer review were reviewed and closed in June 2020 using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations as accepted by NRC in the letter dated May 3, 2017 (References 8.16 and 8.17). There are no open finding level F&Os from the September 2019 peer review. Further, there are no open finding level F&Os for the HNP Internal Events PRA model. Additional finding level F&Os against the Internal Flood model were reviewed and closed in December 2022 using the process documented in NEI 17-07

Relief Request RA-24-0231 Enclosure Page 11 of 32 Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard (Reference 8.18).

There are 4 open findings for the HNP Internal Flood model. Each finding has been resolved or dispositioned for 10 CFR 50.69 implementation and are shown to have no impact or negligible impact on RI-ISI. Application of Code Case N-752 utilizes a very similar analysis to 10 CFR 50.69 and RI-ISI and the PRA quality requirements/impacts for 10 CFR 50.69 and RI-ISI are sufficiently similar to use the same conclusions from 10 CFR 50.69 and RI-ISI for Code Case N-752.

The above information demonstrates that the HNP 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 and is, therefore, acceptable for the ASME Code Case N-752 categorization process.

Robinson The PRA models credited are the same PRA models credited in the following NRC-approved License Amendments:

H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment No. 266 to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors" (Reference 8.19)

H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment No. 265 Regarding Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," (Reference 8.20)

The Robinson Code Case N-752 categorization process for the internal events and flooding hazard uses the plant-specific PRA model.

The Duke Energy risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the Robinson unit.

The PRA 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.

The Robinson internal events PRA model was subject to a self-assessment and a full-scope peer review conducted in May 2010 against the 2009 ASME/ANS standard, RG 1.200 Revision 1, and NEI 05-04. An additional focused scope peer review was completed to

Relief Request RA-24-0231 Enclosure Page 12 of 32 address a model upgrade in 2017 against the 2009 ASME/ANS standard and RG 1.200 Revision 2.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Events PRA model.

The Robinson internal flood PRA model was subject to a self-assessment and full-scope peer review conducted in August 2015 against the 2009 ASME/ANS standard and RG 1.200 Revision 2.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Flooding PRA model.

Closed findings were reviewed and closed in August 2017 for Robinson Internal Events and Internal Flood models using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) as accepted by NRC in the letter dated May 3, 2017.

There are 6 open findings for the Robinson Internal Events PRA model and 4 open findings for the Robinson Internal Flood PRA model. Each finding has been resolved or dispositioned for 10 CFR 50.69 implementation and are shown to have no impact or negligible impact on RI-ISI. Application of Code Case N-752 utilizes a very similar analysis to 10 CFR 50.69 and RI-ISI and the PRA quality requirements/impacts for 10 CFR 50.69 and RI-ISI are sufficiently similar to use the same conclusions from 10 CFR 50.69 and RI-ISI for Code Case N-752.

The above information demonstrates that the Robinson 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 and is, therefore, acceptable for the ASME Code Case N-752 categorization process.

Brunswick The PRA models credited are the same PRA models credited in the following NRC-approved License Amendments:

Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment Nos. 308 and 336 Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-505, Revision 2 (Reference 8.21)

Brunswick Steam Electric Plant, Units 1 And 2 - Issuance of Amendment Nos. 292 and 320 to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors" (Reference 8.22)

Relief Request RA-24-0231 Enclosure Page 13 of 32 Brunswick Steam Electric Plant, Units 1 and 2 -

Issuance of Amendments Regarding request to Relocate Specific Surveillance Frequencies to Licensee Controlled Program (Reference 8.23)

The Brunswick Code Case N-752 categorization process for the internal events and flooding hazard uses the plant-specific PRA model.

The Duke Energy risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the Brunswick unit.

The PRA 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.

The BNP internal events PRA model was subject to a self-assessment and a full-scope peer review conducted in June 2010 against RG 1.200 Revision 2.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Events PRA model.

The BNP internal flood PRA model was subject to a self-assessment and a full-scope peer review conducted in June 2010 and a focused scope peer review covering 28 SRs conducted in December 2016 both against RG 1.200 Revision 2.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Flooding PRA model.

Finding level Facts and Observations (F&Os) related to the BNP internal events and internal flooding PRA models were reviewed and closed in August 2017, December 2019 and May 2020 using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations, as accepted by the NRC in the letter dated May 3, 2017. All Supporting Requirements (SRs) for the BNP internal events and internal flooding PRA models were assessed to be met at least at Capability Category II and there are no open finding level F&Os against either model.

The above information demonstrates that the BNP 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 and is, therefore, acceptable for the ASME Code Case N-752 categorization process.

Relief Request RA-24-0231 Enclosure Page 14 of 32 McGuire The PRA models credited are the same PRA models credited in the following NRC-approved License Amendments:

McGuire Nuclear Station, Units 1 and 2, Issuance of Amendment Nos. 330 and 309, Regarding Revision of Technical Specifications to Adopt Technical Specification Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (Reference 8.25)

McGuire Nuclear Station, Units 1 and 2 - Issuance of Amendment Nos. 331 and 310, Regarding Adoption of Title 10 of the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (Reference 8.24)

McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Revision of the Technical Specifications to Relocate Specific Surveillance Frequencies to a Licensee Controlled Program Using a Risk-Informed Justification (TSTF-425)

(Reference 8.26)

The McGuire Code Case N-752 categorization process for the internal events and flooding hazard uses the plant-specific PRA model.

The Duke Energy risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the McGuire unit.

The PRA 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.

The MNS Units 1 and 2 Internal Events PRA model peer review was performed in June 2015 against the ASME/ANS PRA Standard RA-Sa-2009, RG 1.200 Revision 2, and NEI 05-04.

Resolved findings were reviewed and closed in February 2016 using the process documented in the draft of Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) published at the time of the review. Subsequently, the finding closure review was reperformed in May 2019 to the approved process documented in Appendix X to NEI 05-04/07-12/12-13 as accepted by the NRC in the letter dated May 3, 2017. A subsequent finding closure review was conducted in November 2021 where resolved findings were reviewed and closed using the process documented in NEI 17-07

Relief Request RA-24-0231 Enclosure Page 15 of 32 Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard.

The MNS Units 1 and 2 LERF PRA model peer review was performed in December 2012 against the ASME/ANS RA-Sa-2009, RG 1.200 Revision 2, and NEI 05-04. Resolved findings were reviewed and closed in November 2018 using the process documented in Appendix X to NEI 05-04/07-12/12-13 as accepted by the NRC in the letter dated May 3, 2017. A subsequent finding closure review was conducted in June 2022 where resolved findings were reviewed and closed using the process documented in NEI 17-07.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Events PRA model.

The MNS Units 1 and 2 Internal Flooding PRA model peer review was performed in September 2011 against the ASME/ANS RA-Sa-2009, RG 1.200 Revision 2, and NEI 05-04.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Flooding PRA model.

A finding closure review was conducted on the Internal Flooding PRA model in November 2018 where resolved findings were reviewed and closed using the process documented in Appendix X to NEI 05-04/07-12/12-13 as accepted by the NRC in the letter dated May 3, 2017. A subsequent finding closure review was conducted in June 2022 where resolved findings were reviewed and closed using the process documented in NEI 17-07.

In conclusion, all the finding level F&Os have been closed for the internal events (including LERF) and internal flooding PRA models, and all associated SRs are now judged to be met at Capability Category II or higher. There are no open finding level F&Os against the MNS internal events or internal flooding PRA models.

The above information demonstrates that the MNS 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 and is, therefore, acceptable for the ASME Code Case N-752 categorization process.

Catawba The PRA models credited are the same PRA models credited in the following NRC-approved License Amendments:

Relief Request RA-24-0231 Enclosure Page 16 of 32 Catawba Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Revision of the Technical Specifications to Relocate Specific Surveillance Frequencies to a Licensee-Controlled Program Using a Risk-Informed Justification (TSTF-425) (Reference 8.27)

Catawba Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding National Fire Protection Association Standard NFPA 805 (Reference 8.34)

The Catawba Code Case N-752 categorization process for the internal events and flooding hazard uses the plant-specific PRA model.

The Duke Energy risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for the Catawba unit.

The PRA 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.

The CNS Units 1 and 2 Internal Events PRA model peer review was performed in April 2016 against RG 1.200 Revision 2. Focused scope peer reviews were conducted in September 2017 and December 2021 to address specific model upgrades.

Resolved findings were reviewed and closed in August 2017 using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) as accepted by the NRC in the letter dated May 3, 2017. Subsequent finding closure reviews were conducted in December 2021 and October 2022 where resolved findings were reviewed and closed using the process documented in NEI 17-07.

The CNS Units 1 and 2 LERF PRA model peer review was performed in January 2013 against the 2009 ASME/ANS standard and RG 1.200 Revision 2.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Events PRA model.

Resolved findings were reviewed and closed in November 2018 using the process documented in Appendix X to NEI 05-04/07-12/12-13 as accepted by the NRC in the letter dated May 3, 2017. A subsequent finding closure review was conducted in June 2022 where resolved findings were reviewed and closed using the process documented in NEI 17-07.

Relief Request RA-24-0231 Enclosure Page 17 of 32 The CNS Units 1 and 2 Internal Flooding PRA model peer review was performed in February 2012 against the 2009 ASME/ANS standard, RG 1.200 Revision 2, and NEI 05-04.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Flooding PRA model.

A finding closure review was conducted on the Internal Flooding PRA model in February 2019 where resolved findings were reviewed and closed using the process documented in Appendix X to NEI 05-04/07-12/12-13 as accepted by NRC in the letter dated May 3, 2017.

There are no unreviewed PRA upgrades as defined by RG 1.200 Revision 3 in the Internal Flooding PRA model.

In conclusion, all the finding level F&Os have been closed for the internal events (including LERF) and internal flooding PRA models, and all associated SRs are now judged to be met at Capability Category II or higher. There are no open finding level F&Os against the CNS internal events or internal flooding PRA models.

The above information demonstrates that the CNS 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 and is, therefore, acceptable for the ASME Code Case N-752 categorization process.

D.

Feedback and Process Adjustment Duke Energy 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. Duke Energy 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).

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

Relief Request RA-24-0231 Enclosure Page 18 of 32 treatment is consistent with RISC-3 treatment requirements specified in 10 CFR 50.69(d)(2).

To comply with the above, Duke Energy has already revised existing procedures and will develop/revise documents to define treatment requirements for performing repair/replacement activities on LSS items in accordance with Code Case N-752. Duke Energy defined treatment requirements will address design control, procurement, installation, configuration control, and corrective action.

Duke Energy 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 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.

Relief Request RA-24-0231 Enclosure Page 19 of 32

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 Duke Energy commitments to regulatory and enforcement authorities having jurisdiction at BNP Units 1 and 2, CNS Units 1 and 2, RNP Unit 2, MNS Units 1 and 2, and HNP Unit 1.

11. Periodic ISI and inservice testing (IST) of LSS items at BNP Units 1 and 2, CNS Units 1 and 2, RNP Unit 2, MNS Units 1 and 2, and HNP Unit 1, will continue to be performed as follows:

ISI of LSS pressure-retaining items or their associated supports will be performed in accordance with BNP, CNS, RNP, MNS, HNPs 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 BNP, CNS, RNP, MNS, HNPs 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 BNP, CNS, RNP, MNS, HNPs 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 BNP, CNS, RNP, MNS, HNP.

12.

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 Duke Energys 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).

Relief Request RA-24-0231 Enclosure Page 20 of 32

13.

As permitted by Code Case N-752, Duke Energy intends to implement the exemption on IWA-1400(g) [and (f) depending on Code edition - specific applicable subsections for each site are noted in Table 2] 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, Duke Energy 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).

Table 2 Plant/Unit(s)

ASME Section XI Code Edition/Addenda Exemption to IWA-1400(g) and/or (f)

Brunswick Nuclear Plant (BNP) Units 1 and 2 2007 Edition, Through 2008 Addendum IWA-1400(f)

Catawba Nuclear Station (CNS) Units 1 and 2 2007 Edition, Through 2008 Addendum IWA-1400(f)

H.B. Robinson Nuclear Plant (RNP) Unit 2 2017 Edition IWA-1400(g)

McGuire Nuclear Station (MNS) Unit 1 2019 Edition IWA-1400(g)

McGuire Nuclear Station (MNS) Unit 2 2019 Edition IWA-1400(g)

Shearon Harris Nuclear Plant (HNP) Unit 1 2007 Edition, Through 2008 Addendum IWA-1400(f)

14. As permitted by ASME Code Case N-752, Duke Energy intends to implement the Quality Assurance (QA) Program exemption application to Section XI paragraphs IWA-1400(o)

[and (n) depending on the sites Code Edition - specific applicable subsections for each site are noted in Table 3] and

Relief Request RA-24-0231 Enclosure Page 21 of 32 IWA-4000 when performing repair/replacement activities on SSCs determined to be LSS in accordance with ASME Code Case N-752. However, the exemption from the QA Program requirements of ASME Section XI does not apply if the QA Program requirements of 10 CFR 50, Appendix B, or NQA-1 are required at the facility, as is the case for Duke Energys nuclear facilities. Duke Energys QA Program requirements are described in the Duke Energy Fleet Quality Assurance Program Description (QAPD) whereby proposed changes are subject to the regulatory change control requirements of 10 CFR 50.54(a)(3). To implement the ASME Code Case N-752 exemption from the ASME Section XI QA Program requirements, Duke Energy revised the QAPD in accordance with 10 CFR 50.54(a)(3) to exempt the QAPD requirements for the repair and replacement of Class 2 and 3 components determined to be LSS in accordance with ASME Code Case N-752.

The Basis for the Duke Energy QAPD change is established in the precedent identified as Reference 8.12 of this proposed alternative 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 licensees facility. Within the framework of the amended QAPD, Duke Energy will define alternative treatment requirements that confirm with reasonable confidence that each Class 2 and 3 LSS SSC will remain capable of performing its safety-related function under design-basis conditions. In doing so, Duke Energy will use current QA Program processes and procedures with additional controls for the treatment of Class 2 and 3 LSS components to assure continued capability and reliability of the design-basis function(s). This includes ensuring that changes to the configuration, design, material, fabrication, examination, and testing requirements used to support repair/replacement activities on Class 2 and 3 LSS SSCs are performed in accordance with Duke Energys existing design change process and addressing in Duke Energys corrective action program (CAP) any condition that may prevent an LSS SSC from performing its design-basis function. For the procurement of Class 2 and 3 LSS components as non-safety-related for repair/replacement activities in accordance with ASMEs Code Case N-752, supplemental procurement requirements will be specified, and additional controls will be implemented as appropriate. Such controls include prohibiting suppliers of Class 2 and 3 SSCs and subparts to make design changes or changes to the procurement order without prior Duke Energy approval and conducting receipt inspections using qualified inspection personnel consistent with Duke

Relief Request RA-24-0231 Enclosure Page 22 of 32 Energys procurement requirements. Using these existing QA Program processes and alternative treatment requirements, reasonable confidence exists that the implementation of ASME Code Case N-752 will ensure that each Class 2 and 3 LSS SSC remains capable of performing its design-basis function, and thereby the Duke Energy QAPD will continue to provide an acceptable level of quality and safety.

Table 3 Plant/Unit(s)

ASME Section XI Code Edition/Addenda Exemption to IWA-1400(o) and/or (n)

Brunswick Nuclear Plant (BNP) Units 1 and 2 2007 Edition, Through 2008 Addendum IWA-1400(n)

Catawba Nuclear Station (CNS) Units 1 and 2 2007 Edition, Through 2008 Addendum IWA-1400(n)

H.B. Robinson Nuclear Plant (RNP) Unit 2 2017 Edition IWA-1400(o)

McGuire Nuclear Station (MNS) Unit 1 2019 Edition IWA-1400(o)

McGuire Nuclear Station (MNS) Unit 2 2019 Edition IWA-1400(o)

Shearon Harris Nuclear Plant (HNP) Unit 1 2007 Edition, Through 2008 Addendum IWA-1400(n)

15. As permitted by Code Case N-752, Duke Energy intends to implement the exemptions on IWA-1400(k) [and (j) depending on Code edition - specific applicable subsections for each site are noted in Table 4] and IWA-4000 applicable to repair/replacement programs and plans. In lieu of these ASME Section XI administrative controls, Duke Energy will establish Owner-defined administrative controls as required by paragraph

-1420(a) of Code Case N-752. Duke Energy 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 (j) depending on Code edition

- specific applicable subsections for each site are noted in

Relief Request RA-24-0231 Enclosure Page 23 of 32 Table 4] 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).

Table 4 Plant/Unit(s)

ASME Section XI Code Edition/Addenda Exemption to IWA-1400 (k) and/or (j)

Brunswick Nuclear Plant (BNP) Units 1 and 2 2007 Edition, Through 2008 Addendum IWA-1400(j)

Catawba Nuclear Station (CNS) Units 1 and 2 2007 Edition, Through 2008 Addendum IWA-1400(j)

H.B. Robinson Nuclear Plant (RNP) Unit 2 2017 Edition IWA-1400(k)

McGuire Nuclear Station (MNS) Unit 1 2019 Edition IWA-1400(k)

McGuire Nuclear Station (MNS) Unit 2 2019 Edition IWA-1400(k)

Shearon Harris Nuclear Plant (HNP) Unit 1 2007 Edition, Through 2008 Addendum IWA-1400(j)

16.

As permitted by Code Case N-752, Duke Energy 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, Duke Energy 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 Duke Energy program will utilize

Relief Request RA-24-0231 Enclosure Page 24 of 32 existing Duke Energy 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.

Duke Energy 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).

17.

As permitted by Code Case N-752, Duke Energy intends to implement the documentation exemptions on IWA-6211 [6210 depending on Code edition - specific applicable sections for each site are noted in Table 5] (d), IWA-6211 [6210 depending on Code edition - specific applicable sections for each site are noted in Table 5] (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 Duke Energys 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.

Relief Request RA-24-0231 Enclosure Page 25 of 32 Table 5 Plant/Unit(s)

ASME Section XI Code Edition/Addenda Exemption to IWA-6211 (d) and (e),

and/or IWA-6210 (d) and (e)

Brunswick Nuclear Plant (BNP) Units 1 and 2 2007 Edition, Through 2008 Addendum IWA-6210(d) and (e)

Catawba Nuclear Station (CNS) Units 1 and 2 2007 Edition, Through 2008 Addendum IWA-6210(d) and (e)

H.B. Robinson Nuclear Plant (RNP) Unit 2 2017 Edition IWA-6211(d) and (e)

McGuire Nuclear Station (MNS) Unit 1 2019 Edition IWA-6211(d) and (e)

McGuire Nuclear Station (MNS) Unit 2 2019 Edition IWA-6211(d) and (e)

Shearon Harris Nuclear Plant (HNP) Unit 1 2007 Edition, Through 2008 Addendum IWA-6210(d) and (e)

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 8.10). Repair/replacement activities performed on items determined to be HSS 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.

Duke Energys proposed treatment requirements, as described herein, meet these criteria, and provide reasonable confidence that LSS systems and items remains 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

Relief Request RA-24-0231 Enclosure Page 26 of 32 activities are consistent with NRC requirements specified in 10 CFR 50.69.

6.0 DURATION OF PROPOSED ALTERNATIVE The duration of this relief request is for the remainder of the current operating licenses for BNP Units 1 and 2, CNS Units 1 and 2, RNP Unit 2, MNS Units 1 and 2, and HNP Unit 1, as shown below.

Table 6 BNP Current Operating Licenses Docket Number License Expires Unit 1 50-325 09/08/2036 Unit 2 50-324 12/27/2034 Table 7 CNS Current Operating Licenses Docket Number License Expires Unit 1 50-413 12/05/2043 Unit 2 50-414 12/05/2043 Table 8 RNP Current Operating License Docket Number License Expires Unit 2 50-261 07/31/2030 Table 9 MNS Current Operating Licenses Docket Number License Expires Unit 1 50-369 06/12/2041 Unit 2 50-370 03/03/2043

Relief Request RA-24-0231 Enclosure Page 27 of 32 Table 10 HNP Current Operating License Docket Number License Expires Unit 1 50-400 10/24/2046 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 (ML20148M343), approved May 19, 2021 (ML21118B039).

7.2. Duke Energy, Oconee Nuclear Station Units 1, 2, and 3 Request for Relief RA 0174, submitted July 27, 2022 (ML22208A031),

approved December 13, 2023 (ML23262A967).

7.3. Entergy Operations, Inc., Grand Gulf Nuclear Station, Unit 1; River Bend Station, Unit 1; Waterford Steam Electric Station, Unit 3 Request for Relief No. EN-RR-22-001, submitted June 30, 2022 (ML22181B114), approved May 30, 2024 (ML24060A219).

7.4. NextEra Energy Point Beach, LLC/NextEra Energy Seabrook, LLC/Florida Power & Light Company, Saint Lucie Nuclear Plant, Units 1 And 2; Turkey Point Nuclear Plant, Units 3 And 4; Seabrook Nuclear Plant; And Point Beach Nuclear Plant, Units 1 And 2 Fleet Relief Request (FRR) 23-01, submitted March 15, 2023 (ML23074A155), approved June 12, 2024 (ML24149A286).

7.5. Several domestic nuclear power plants have sought and obtained approval to apply the risk-informed evaluation and categorization (classification) process of Relief Request ANO2-R&R-004, Revision 1, for repair/replacement activities for Class 2 and Class 3 pressure-retaining items or their associated supports. These include the following Duke Energy Sites:

7.5.1. NRC Letter, Brunswick Steam Electric Plant, Units 1 and 2 -

Issuance of Amendment Nos. 305 and 333 to revise License Conditions to Modify Approved 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, dated April 30, 2021 (ML21067A224).

Relief Request RA-24-0231 Enclosure Page 28 of 32 7.5.2. NRC letter to Duke Energy, H.B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment No. 266 to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, dated September 24, 2019 (ML19205A289).

7.5.3. NRC letter to Duke Energy, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment No. 174 RE: Adopt Title 10 of the Code of Federal Regulations 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, dated September 17, 2019 (ML19192A012).

8.0 REFERENCES

8.1. American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, 2007 Edition with the 2008 Addenda.

8.2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, 2017 Edition.

8.3. American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, 2019 Edition.

8.4. RIS 2004-12, Clarification on use of Later Editions and Addenda to the ASME OM Code and Section XI (ML042090436).

8.5. Brunswick Steam Electric Plant, Units 1 and 2; Catawba Nuclear Station, Units 1 and 2; H. B. Robinson Steam Electric Plant, Unit 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3; and Shearon Harris Nuclear Power Plant, Unit 1 - Request RA 0262, submitted September 16, 2020 (ML20260H325), approved November 9, 2020 (ML20300A206).

8.6. RIS 2004-16, Use of Later Editions and Addenda to ASME Code Section XI for Repair/Replacement Activities (ML042590067).

8.7. Brunswick Steam Electric Plant, Units 1 and 2; Catawba Nuclear Station, Units 1 and 2; H. B. Robinson Steam Electric Plant, Unit 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3; and Shearon Harris Nuclear Power Plant, Unit 1 - Request RA-20-0263, submitted September 16, 2020 (ML20260H326), approved May 6, 2021 (ML21113A013).

8.8. 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, And Components for Nuclear Power Reactors,"

USNRC, 69 FR 68047, Nov. 22, 2004.

Relief Request RA-24-0231 Enclosure Page 29 of 32 8.9. 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," dated May 2006.

8.10. 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," (ML071150108) as supplemented by letters dated August 6, 2007 (ML072220160), February 20, 2008 (ML080520186), and January 12, 2009 (ML090120620).

8.11. Safety Evaluation Report (SER) 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," dated April 22, 2009 (ML090930246).

8.12. NRC Letter to Entergy, Arkansas Nuclear One, Units 1 and 2 - Request for Approval of Change to the Entergy Quality Assurance Program Manual, dated May 19, 2021 (ML21132A279).

8.13. Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment No. 174 Re: Adopt Title 10 of the Code of Federal Regulations 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors," dated September 17, 2019, ADAMS Accession No. ML19192A012.

8.14. Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment No. 184 Regarding Technical Specifications Task Force (TSTF)

Traveler TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF Initiative 4B, dated April 2, 2021, Adams Accession No. ML21047A314.

8.14.1. Shearon Harris Nuclear Power Plant, Unit 1 - Correction to Amendment No. 184 Regarding Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF Initiative 4B dated May 4, 2021, Adams Accession No. ML21116A207 8.15. Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Regarding Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program, dated November 29, 2016, Adams Accession No. ML16200A285.

8.16. NEI Letter to NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os), February 21, 2017, (ADAMS Accession Number ML17086A431).

Relief Request RA-24-0231 Enclosure Page 30 of 32 8.17. NRC Letter to Mr. Greg Krueger (NEI), U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os), May 3, 2017, (ADAMS Accession Number ML17079A427).

8.18. Nuclear Energy Institute (NEI) 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, (ADAMS Accession No. ML19241A615).

8.19. H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment No. 266 to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors," dated September 24, 2019, Adams Accession No. ML19205A289.

8.20. H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Amendment No. 265 Regarding Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b,"

dated August 15, 2019, Adams Accession No. ML19158A307.

8.21. Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment Nos. 308 and 336 Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-505, Revision 2, dated May 2, 2022, Adams Accession No. ML22082A268.

8.22. Brunswick Steam Electric Plant, Units 1 And 2 - Issuance of Amendment Nos.

292 and 320 to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors," dated September 17, 2019, Adams Accession No. ML19149A471.

8.23. Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendments Regarding request to Relocate Specific Surveillance Frequencies to Licensee Controlled Program, dated May 24, 2017 (ADAMS Accession No. ML17096A129).

8.24. McGuire Nuclear Station, Units 1 and 2 - Issuance of Amendment Nos.

331 and 310, Regarding Adoption of Title 10 of the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated April 8, 2024, Adams Accession No. 24052A306.

8.25. McGuire Nuclear Station, Units 1 and 2, Issuance of Amendment Nos.

330 and 309, Regarding Revision of Technical Specifications to Adopt Technical Specification Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -

RITSTF Initiative 4b, dated March 26, 2024, Adams Accession No.

24031A540.

Relief Request RA-24-0231 Enclosure Page 31 of 32 8.26. McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Revision of the Technical Specifications to Relocate Specific Surveillance Frequencies to a Licensee Controlled Program Using a Risk-Informed Justification (TSTF-425), dated March 29, 2011, Adams Accession No. ML110680357.

8.27. Catawba Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Revision of the Technical Specifications to Relocate Specific Surveillance Frequencies to a Licensee-Controlled Program Using a Risk-Informed Justification (TSTF-425), dated March 28, 2011, Adams Accession No. ML110670536.

8.28. NEI 00-02 Probabilistic Risk Assessment (PRA) Peer Review Process Guidance January 2000.

8.29. ASME RA-Sb-2005, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum B to ASME RA-S-2002, ASME, New York, New York, December 30, 2005.

8.30. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 1, US Nuclear Regulatory Commission, January 2007.

8.31. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, US Nuclear Regulatory Commission, March 2009.

8.32. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3, US Nuclear Regulatory Commission, December 2020.

8.33. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, ANS, La Grange Park, Illinois, February 2009.

8.34. Catawba Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding National Fire Protection Association Standard NFPA 805, dated June 19, 2017, Adams Accession No. ML17144A034.

8.35. Duke Energy, McGuire Nuclear Station Units 1 and 2, Request for Relief RA 0031, submitted January 23, 2020 (ML20023A272),

approved August 21, 2020 (ML20230A205).

Relief Request RA-24-0231 Enclosure Page 32 of 32 8.36. Brunswick Steam Electric Plant, Units 1 and 2; Catawba Nuclear Station, Units 1 and 2; H. B. Robinson Steam Electric Plant, Unit 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3; and Shearon Harris Nuclear Power Plant, Unit 1 - Request RA-20-0191, submitted September 21, 2020 (ML20265A028),

approved February 16, 2021 (ML21029A335).

8.37. Brunswick Steam Electric Plant, Units 1 and 2; Catawba Nuclear Station, Units 1 and 2; H. B. Robinson Steam Electric Plant, Unit 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3; and Shearon Harris Nuclear Power Plant, Unit 1 - Request RA-23-0001, submitted February 2, 2023 (ML23033A037), approved May 1, 2023 (ML23118A076).