ML23331A921

Public - Non-Concurrence of SE for Proposed Alternative RR-22-0174 for Oconee Nuclear Station, Units 1, 2, and 3 Duke Energy Carolinas, LLC

ML23331A921
Person / Time
Site:	Oconee
Issue date:	11/16/2023
From:	Ballard B Division of Operating Reactor Licensing
To:
References
Download: ML23331A921 (1)

Text

SECTION A: NON-CONCURRENCE OF SAFETY EVALUATION FOR PROPOSED ALTERNATIVE RR-22-0174 FOR OCONEE NUCLEAR STATION, UNITS 1, 2 and 3 DUKE ENERGY CAROLINAS, LLC NON-CONCURRENCE PROCESS TRACKING NUMBER NCP-2023-005 The non-concurring staff (NC staff) provide the following basis for our position on the Division of Operating Licensing (DORL) proposed safety evaluation (ML23262A967) to address the letter dated July 27, 2022, as supplemented by letters dated March 9 and October 20, 2023, in which Duke Energy Carolinas, LLC (Duke Energy, the licensee) submitted alternative request RR-22-0174 for Oconee Nuclear Station (ONS), Units 1, 2, and 3 to the U.S. Nuclear Regulatory Commission (NRC). The NC staffs concerns are that issuance of the current DORL safety evaluation through Title 10 of the Code of Federal Regulations (10 CFR) 50.55a Codes and Standards, under 10 CFR 50.55a(z)(1), would allow a licensee to expect exemption of 10 CFR Part 50, Appendix B quality assurance (QA) requirements and allow an unacceptable level of quality and safety in association with the alternative treatment options for repair/replacement activities of safety-related Low Safety Significant (LSS) items at ONS. Further, in order to address the differences in opinion of the legal requirements to evaluate Duke Energys request, the NC staff propose a set of questions for the Office of the General Counsel (OGC).

Each of the NC staff reviewed Duke Energys request RR-22-0174 to implement the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME) Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1, to determine the risk-informed categorization of ASME Class 2 and 3 items, and to implement alternative treatment for repair/replacement activities for safety-related LSS items in lieu of certain ASME Code,Section XI, paragraph IWA-1000, IWA-4000, and IWA-6000 requirements. As part of this review, the NC staff considered the operating experience of the use of ASME Code Case N-752 in the review for the plant-specific application at ONS.

Issue 1 Duke Energy has requested authorization of their proposed alternative to use ASME Code Case N-752 without exception or deviation. Under these conditions, the NC staff's understanding is for Duke Energy to implement their proposed alternative of Section 5.2.E.14, As permitted by Code Case N-752, Duke Energy intends to implement the QA Program exemption applicable to IWA-1400(n), at ONS, they would need an exemption from 10 CFR Part 50 Appendix B for those applicable LSS items. Additionally, Duke Energy uses the term reasonable confidence to describe the level of the capability of safety-related items to meet their safety function.

Neither the licensee nor ASME Code Case N-752 defines reasonable confidence. Based on the development of ASME Code Case N-752, the NC staff understands that the term reasonable confidence is only explained in the Federal Register Notice for 10 CFR 50.69.

Under 10 CFR 50.69, for a licensee to apply the standard of reasonable confidence, a licensee is expressly exempt from the requirements of 10 CFR Part 50, Appendix B, only upon issuance Section A: Attachment

2 of the license amendment to apply 10 CFR 50.69 at a plant. As a further complication, the NC staff believe that Entergy Operations, Inc. (Entergy) has assumed that they received just such an exemption based on its previous authorization to use ASME Code Case N-752 and the QA manual change at Arkansas Nuclear One, Units 1 and 2 (ANO). The NC staff believe this assumption by Entergy is made clear by statements in several licensee documents as shown in Enclosure I. Through the audit process, the NC staff are concerned that Duke Energy will utilize 10 CFR 50.54(a)(3)(ii), as a basis, to claim this same perceived exemption as Entergy believes they have obtained. The NC staff notes that neither Entergy nor Duke Energy has, or is, requesting explicit exemption from Appendix B under 10 CFR 50.12 for those applicable ASME Code Case N-752 items at ANO or ONS, respectively. The NC staff believe such an exemption is necessary prior to implementing the proposed alternative or Duke Energy should modify their proposed alternative. If neither option is employed, the NC staff are concerned that regulatory uncertainty would result from the inconsistency between the understanding of the applicability of 10 CFR Part 50, Appendix B, by the NRC staff and licensees. For details and specific statements by Entergy and Duke Energy that have led to this safety concern, see Enclosure I below.

Issue 2 The NC staff identified a concern related to the quality of the alternative treatment options in the Duke Energy proposal in lieu of the current regulatory requirements for repair/replacement activities. To address this concern, the NC staffs review of the specific language in the licensees proposed alternative identified several areas of uncertainty in providing reasonable assurance of meeting nationally approved codes and standards. This was the basis for audit and request for additional information (RAI) questions to Duke Energy requesting the licensee to identify the specific nationally approved codes and standards for their proposed alternative. The NC staff found Duke Energys response to not provide specific codes and standards led to further uncertainty in assessing the safety margins and effectiveness of defense-in-depth measures. Therefore, the NC staff cannot find that the licensees proposed alternative provides for an adequate level of quality and safety. In Enclosure II, the NC staff provide their basis for denial of the licensees proposed alternative request. The NC staff continue to believe that this issue could be addressed if the licensee established minimum requirements for specific codes and standards for repair/replacement activities associated with the implementation of the plant-specific proposed alternative at ONS.

Issue 3 For both of the above issues, the NC staff believe that OGC review of the safety evaluation with consideration of specific questions, listed below, would help address the difference in opinions between the NC staff and other NRC staff and management on these issues.

As allowed by Section 5 in LIC-102, Revision 3, Review of Relief Requests, Proposed Alternatives, and Requests to Use Later Code Editions and Addenda, the technical branch chief and subject matter experts can request OGC legal review of relief/alternative requests on the basis of perceived unique or special circumstances. This is a unique review with special circumstances that was initially brought to the attention of the NRC staff by OGC during the 10 CFR 50.55a rulemaking for incorporating ASME Code Cases, including ASME Code Case N-752, into 10 CFR 50.55a. The OGC comment on ASME Code Case N-752 included whether the use of the Code Case by a licensee deviates from the Commission direction that a license amendment is needed to apply risk-informed categorization per 10 CFR 50.69. This is due to

3 the fact that the technical basis for the licensees safety categorization process to utilize ASME Code Case N-752 is based on it being the same as the passive categorization method of 10 CFR 50.69. Note that the proposed alternative states, The categorization and treatment requirements of Code Case N-752 are consistent with those in 10 CFR 50.69. The NRC regulations in 10 CFR 50.69 allow removal of certain special treatment requirements, such as 10 CFR Part 50, Appendix B for repair/replacement and quality control for safety-related LSS structures, systems, and components (SSCs). This also includes the policy issue of whether approving alternative requests for use of ASME Code Case N-752 deviates from the Commission direction that a license amendment is needed to apply risk-informed categorization per 10 CFR 50.69. Therefore, the DORL safety evaluation should be reviewed by OGC to determine if this contradicts the Commission's policy, which would also include OGC assistance in resolving the following questions related to the request by Duke Energy under 10 CFR 50.55a(z)(1) to use ASME Code Case N-752 at ONS:

1. Confirm that the Staff cannot authorize the use of ASME Code Case N-752 (which the licensee states is consistent with 10 CFR 50.69) under the 10 CFR 50.55a(z) alternative process because it bypasses 10 CFR 50.69(b)(2) in which a license amendment is required. In other words, confirm that recategorization of SSCs to high-safety significant (HSS)/LSS and subsequent changes in special treatment of safety-related SSCs is only permitted under the auspice of 10 CFR 50.69.

2. ASME Code Cases within 10 CFR 50.55a do not allow a licensee to be exempt from NRC special treatment requirements in other portions of the NRC regulations, such as the QA requirements in 10 CFR Part 50, Appendix B. Confirm that 10 CFR 50.55a and 10 CFR 50.54 does not allow a licensee to be exempt from the QA requirements in 10 CFR Part 50, Appendix B.

3. Confirm that the previous plant-specific authorization under 10 CFR 50.55a(z) to use ASME Code Case N-752 for ANO does not constitute a past precedent that would require the evaluation of backfit issues for the use of ASME Code Case N-752 at ONS.

For additional basis and background on these questions, please see Enclosure III.

The NC staff have concerns regarding the generic application of ASME Code Case N-752.

However, that discussion is not included as part of this NC.

ENCLOSURE I.

10 CFR PART 50, APPENDIX B APPLICABILITY The Division of Operating Reactor Licensing (DORL) in the NRC Office of Nuclear Reactor Regulation (NRR) has prepared a safety evaluation (SE) authorizing Duke Energy Carolinas, LLC (Duke Energy) to use American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME) Code Case N-752 at Oconee Nuclear Station (ONS) under Section 50.55a(z)(1) in Title 10 of the Code of Federal Regulations (10 CFR 50.55a(z)(1)) as providing an acceptable level of quality and safety. The DORL SE (if issued) might inadvertently result in Duke Energy believing that the NRC staff has authorized an exemption from the Quality Assurance (QA) requirements in 10 CFR Part 50, Appendix B without meeting 10 CFR 50.12, Exemptions, or 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power plants, when implementing ASME Code Case N-752 at ONS.

DORL issued a similar SE on May 19, 2021, authorizing Entergy Operations, Inc. (Entergy) to use ASME Code Case N-752 at Arkansas Nuclear One, Units 1 and 2 (ANO) under 10 CFR 50.55a(z)(1). Entergy has interpreted this SE as allowing ANO to not meet the QA requirements of 10 CFR Part 50, Appendix B for safety-related LSS components when implementing ASME Code Case N-752 without an exemption under 10 CFR 50.12. Entergy had not submitted a request to implement 10 CFR 50.69 at ANO until May 26, 2021, which is later than the NRC staff review and authorization of the Entergy request to implement Code Case N-752. Further, Duke Energy has not submitted requests for exemptions under 10 CFR 50.12 or to implement 10 CFR 50.69 at ONS, which might have been used to justify a reduced level of QA for safety-related LSS components. Therefore, the NRC regulations in 10 CFR 50.12 and 50.69 are not applicable to the request to implement Code Case N-752 at ANO or ONS.

During preparation of the SE by DORL, the NRR Mechanical Engineering and Inservice Testing Branch (EMIB) provided proposed SE input with a request for additional information (RAI) to DORL to clarify that an exemption to 10 CFR Part 50, Appendix B is not authorized for licensees implementing ASME Code Case N-752. Duke Energy and Entergy submittals for ONS and ANO, respectively, requesting to use Code Case N-752 under 10 CFR 50.55a(z)(1), improperly indicate that 10 CFR Part 50, Appendix B will not be met for safety-related LSS components when implementing the Code Case. In addition, statements by NRC staff members regarding the applicability of 10 CFR Part 50, Appendix B, to licensees applying Code Case N-752 have not been consistent.

The proposed SE input by EMIB with the RAI was as follows in italics:

In the proposed alternative to implement ASME Code Case N-752 (dated July 27, 2022) in accordance with 10 CFR 50.55a(z)(1) and during audit discussions regarding the alternative request, the NRC staff found that Duke Energy, the licensee for Oconee, is not correct regarding the applicability of 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, when implementing ASME Code Case N-752 at the Oconee nuclear power plant. An alternative request under the NRC regulations in 10 CFR 50.55a(z) cannot authorize an exemption from the quality assurance (QA) requirements in 10 CFR Part 50, Appendix B, at a nuclear power plant. An exemption from 10 CFR Part 50, Appendix B, can only be obtained through a request by a licensee under 10 CFR 50.12, Exemptions, or a license amendment request under 50.90 of the NRC regulations. One precedent would be a license amendment request under 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components

I.3 ASME Code Case N-752 states that the Owner shall define requirements to confirm with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design-basis conditions. Code Case N-752 does not define the term reasonable confidence with respect to the QA requirements in 10 CFR Part 50, Appendix B. The use of the term reasonable confidence is introduced by the NRC regulations as part of 10 CFR 50.69 for the treatment of safety-related SSCs that are exempted from Appendix B due to their low safety significance, in accordance with the provisions of 10 CFR 50.69.

Reasonable confidence is treatment that is less than the treatment mandated by special treatment requirements (such as 10 CFR Part 50, Appendix B). For example, the Commission stated in the Federal Register notice 69 FR 68008 (dated November 22, 2004) at 68041 the following:

By reasonable confidence, the Commission means that the licensee or applicant is required to provide a reasonable confidence level with regard to maintaining the capability of RISC [Risk-Informed Safety Class]-3 safety-related functions. As indicated previously in this notice, reasonable confidence is a level of confidence that is both less than that associated with RISC-1 SSCs which are subject to all the special treatment requirements, and consistent with their individual low safety significance. The term ensure is intended to convey the Commissions determination that the licensee is under a legally-binding regulatory requirement to provide the requisite reasonable confidence.

The Oconee licensee has not defined the term reasonable confidence for its implementation of ASME Code Case N-752 with respect to the QA requirements in 10 CFR Part 50, Appendix B, which still applies to the safety-related components at Oconee (50.69 has not been approved at Oconee) regardless of the safety classification under N-752.

Instead of applying the 50.69 definition of reasonable confidence to the current application, the staff has evaluated the proposed changes in treatment on their merits, as discussed below.

In its alternative request dated July 27, 2022, Duke Energy indicated that, in accordance with the process allowed by 10 CFR 50.54(a)(3)(ii), Duke Energy will not request prior NRC approval of a planned change to the Oconee QAPD because the NRC has already approved a similar change for the ANO application of N-752, and Duke Energy can show that its QAPD changes do not reduce effectiveness compared to the changes to the ANO QAPD. Duke Energy references the NRC safety evaluation (SE) for the ANO QAPD[Quality Assurance Procedure Manual] change associated with N-752, dated May 19, 2021 (ML211132A279). That SE concludes that there is reasonable assurance that the specific changes to the ANO QAPM [QA Program Manual] will continue to meet the requirements of Appendix B for the treatment of safety-related SSCs identified as LSS while implementing Code Case N-752.

In its response to an NRC request for additional information (RAI) dated March 9, 2023, Duke Energy stated that it is requesting to use ASME Code Case N-752 with no exceptions or deviations. The NRC staff considers that Duke Energy is supplementing Code Case N-752 such that this statement in the RAI response is not correct.

In its alternative request dated July 27, 2022, the Oconee licensee relies on the NRC staff review of an alternative request submitted by Entergy to apply ASME Code Case N-752 at

I.4 the ANO nuclear power plant. Despite incorrect statements1 by Entergy regarding the applicability of 10 CFR Part 50, Appendix B, when implementing ASME Code Case N-752 at ANO, the NRC staff determined that the specific treatment controls for safety-related LSS components will meet the QA requirements in accordance with 10 CFR Part 50, Appendix B, while allowing more flexibility to Entergy for the procurement of such LSS components at ANO. The NRC issued an SE dated May 19, 2021 (ML21118B039) authorizing the use of ASME Code Case N-752 at the ANO nuclear power plant, and another SE dated May 19, 2021 (ML21132A279) approving the proposed change to the QAPM with the specified procurement controls for safety-related LSS components at the ANO nuclear power plant.

The NRC staff cross-referenced its SE authorizing the use of ASME Code Case N-752 at the ANO nuclear power plant with acceptance of the specified procurement controls for safety-related LSS components in its SE approving the proposed change to the QAPM for the ANO nuclear power plant. Therefore, the NRC staff authorization of the use of ASME Code Case N-752 is based on the implementation of the procurement controls for safety-related LSS components specified by the ANO licensee.

The NRC regulations in 10 CFR 50.54(a) allow a licensee to modify its description of its QA program without prior NRC review, provided the QA requirements in 10 CFR Part 50, Appendix B, will continue to be met. In its alternative request, Duke Energy indicates that it will implement the procurement controls accepted by the NRC staff as proposed by Entergy for implementation of ASME Code Case N-752 at ANO. Therefore, the staff finds it acceptable for Duke Energy to adopt similar QAPD changes to those approved for ANO, subject to the provisions of 10 CFR 50.54.

RAI: An alternative request under 10 CFR 50.55a(z) cannot waive 10 CFR Part 50, Appendix B, requirements; therefore, the NRC staff cannot authorize the subject request as written. The NRC staff approved the ANO precedent, referenced in the request, on the basis that the QAPD changes were acceptable for meeting Appendix B for items characterized as LSS under CC N-752. The staff determined that ANOs relaxed requirements that did not constitute a reduction in effectiveness for the LSS items and continue to meet Appendix B. Based on the above, please clarify the statements regarding (1) the applicability of Appendix B, which will apply during the use of CC N-752 at Oconee, (2) the use of CC N-752 without exception to include Note 1, because Appendix B will continue to apply at Oconee, (3) meeting reasonable confidence vs.

reasonable assurance, to confirm that Appendix B will continue to be met at Oconee, and (4) updates to the Duke QAPD regarding Class 2 and 3 LSS items in that these items are not exempt from Appendix B. Also, clarify whether the above results in changes to Dukes plans for updating the QAPD under the provision of 10 CFR 50.54.

Based on the RAI response by the licensee dated XX,XX, 2023, the NRC staff finds that the Oconee licensee has corrected its compliance with the QA requirements of 10 CFR Part 50, Appendix B, to support the NRC staff authorization of the use of ASME Code Case N-752 at the Oconee nuclear power plant to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

1 The NRC staff notes that a submittal dated October 26, 2020 (ML20300A324) by the ANO licensee incorrectly stated that the QA requirements in 10 CFR Part 50, Appendix B, would not be met for safety-related LSS components when implementing ASME Code Case N-752 at the ANO nuclear power plant.

I.6 (4) Item 14 in the Duke Energy submittal dated July 27, 2022 (ML22208A031) states the following:

As permitted by Code Case N-752, Duke Energy intends to implement the QA Program exemption applicable to IWA-1400(n) 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, Duke Energy will update the Fleet Quality Assurance Program Description (QAPD) 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 QAPD. Duke Energy will develop elements describing treatment of these LSS SSCs to ensure continued capability and reliability of the design basis function. In accordance with 10 CFR 50.54(a)(3)(ii), Duke Energy is not requesting prior NRC approval of the change to the QAPD because it has previously been approved for Entergy (Reference 8.14) in conjunction with a request for Arkansas Nuclear One to adopt ASME Code Case N-752 (References 8.15 and 8.16).

In Item 14, Duke Energy is relying on the Entergy submittals which, as noted above, specify that Appendix B will not be met for safety-related LSS components when implementing ASME Code Case N-752 at ANO.

(5) Duke Energy submittal dated March 9, 2023 (ML23068A015) states in response to RAI 1.a the following:

Duke Energy is requesting to use ASME Code Case N-752 with no exceptions or deviations.

Duke Energy submittal dated March 9, 2023 (ML23068A015) states in response to RAI 5.b the following:

Code Case N-752 is limited to Class 2 and 3 items. All unanalyzed Class 2 and 3 components will continue to meet their applicable nuclear special treatment requirements (e.g., Repair & Replacement per ASME Section XI requirements, QA per Appendix B, etc.).

In that Duke Energy states that it is requesting the use of ASME Code Case N-752 without exception or deviation, the NRC SE cannot be used to require a condition on the use of ASME Code Case N-752 at ONS. Further, Duke Energy states that unanalyzed Class 2 and 3 components will meet Appendix B, which indicates that the analyzed Class 2 and 3 components will not meet Appendix B.

(6) On September 29, 2023, the NRR Project Manager for ANO forwarded the Entergy submittal to modify its Safety Analysis Report (SAR) to reflect the NRC authorization to use ASME Code Case N-752 at ANO. The modified SAR for ANO states the following:

Code Case N-752 provides a process for determining the risk-informed categorization and treatment for repair/replacement activities on pressure retaining Class 2 and 3 components and their associated supports.

I.7 Components are categorized as either High Safety Significant (HSS) or Low Safety Significant (LSS).

Repair/replacement activities on Class 2 and 3 pressure retaining components and supports categorized as HSS shall continue to comply with the ASME XI Code and Entergy QAPM. Alternatively, Class 2 and 3 pressure retaining components and supports determined to be LSS in accordance with Code Case N-752 may comply with the alternative treatment requirements of Code Case N-752 including those specified below.

1. Compliance with the repair/replacement requirements of ASME Section XI (e.g., IWA-4000) is not required.

2. Compliance with the Entergy QAPM is not required.

In implementing Code Case N-752, the fracture toughness requirements specified in Owners Requirements and the original construction Code applicable to LSS components shall be met. Additionally, Inservice Inspections (ISI) and Inservice Testing (IST) of LSS components shall continue to be performed in accordance with the sites ISI and IST programs.

Class 2 and 3 pressure retaining components and supports that have been categorized as LSS in accordance with Code Case N-752 are identified as LSS in the Enterprise Asset Management (EAM) application. Alternative treatment requirements for LSS components and supports are specified in applicable risk-informed repair/replacement program procedures.

The SAR update for ANO submitted by Entergy does not reference the two NRC safety evaluations (both dated May 19, 2021, at ML21132A279 and ML21118B039) that authorized the use of ASME Code Case N-752 at ANO under 10 CFR 50.55a(z)(1),

based on the specific QA controls for safety-related LSS components described in the ANO alternative request and its supplemental submittals (such as dated April 5, 2021, and April 30, 2021). The SAR update also does not indicate that the NRC SE on the QA controls for safety-related LSS components (ML21132A279) stated that the NRC staff concludes that there is reasonable assurance that the licensees QAPM will continue to meet the requirements of Appendix B to 10 CFR Part 50 while implementing ASME Code Case N-752 for the treatment of safety-related SSCs identified as LSS.

In addition to these statements in ASME Code Case N-752, and Duke Energy and Entergy documents, industry personnel have indicated that 10 CFR Part 50, Appendix B does not apply to safety-related LSS components when implementing Code Case N-752. Further, recent industry presentations have also failed to indicate that 10 CFR Part 50, Appendix B continues to apply for safety-related LSS components when implementing ASME Code Case N-752.

On October 11, 2023, DORL distributed for concurrence its final SE to authorize Duke Energy to implement ASME Code Case N-752 at ONS under 10 CFR 50.55a(z)(1) as providing an acceptable level of quality and safety. Rather than including the EMIB SE input, DORL stated that the input provided by the NRR Division of Reactor Oversight, Quality Assurance and Vendor Inspection Branch (IQVB) would be used for the SE despite the fact that the IQVB input does not resolve inconsistent licensee submittals and statements. Further, the DORL SE does not include EMIB in the concurrence chain although EMIB provided significant participation in

I.8 the NRC staff review of the Duke Energy request to use Code Case N-752 at ONS. Also, the DORL SE includes numerous references to licensee documents dated October XX that have not been submitted to the NRC.

NRR Office Instruction LIC-102, Revision 3, Review of Relief Requests, Proposed Alternatives, and Requests to Use Later Code Editions and Addenda, states the following on pages 9 and 10:

OGC OGC legal review of reliefs/alternatives is not required. However, an NRR stakeholder may suggest, on the basis of perceived unique or special circumstances, that a relief/alternative be reviewed by OGC. When this happens, the determination of need for OGC review should be jointly made by the technical branch chief, licensing branch chief, and subject matter expert for relief requests/proposed alternatives.

Based on the above facts, the DORL SE (if issued) authorizing the use of ASME Code Case N-752 at ONS under 10 CFR 50.55a(z)(1) will violate the NRC rules for granting exemptions from the NRC regulations in light of information in ASME Code Case N-752, and Duke Energy and Entergy documents, indicating that 10 CFR Part 50, Appendix B will not be met for safety-related LSS components when implementing the Code Case. In the past, OGC has indicated that an SE cannot impose conditions when authorizing a 10 CFR 50.55a(z) alternative request. Based on the significance of the NRC decision, the Duke Energy request to apply ASME Code Case N-752 as an alternative under 10 CFR 50.55a(z)(1) should be reviewed by OGC. Following resolution of the DORL SE for the use of Code Case N-752 at ONS, the NRC needs to resolve the misinterpretation by Entergy that the SE issued to allow the use of ASME Code Case N-752 at ANO includes an exemption from 10 CFR Part 50, Appendix B for safety-related LSS components.

ENCLOSURE II.

ALTERNATIVE CODES AND STANDARDS CONCERN PROPOSED SAFETY EVALUATION DENIAL Enclosure II provides the non-concurring (NC) staffs basis for denial of the licensees proposed alternative treatment process for RR-22-0174. Given the open-ended allowances provided by the licensee in Sections 5.2.E.6, 5.2.E.7 and 5.2.E.8, of their submittal, for the use of national codes and standards for repair/replacement activities, nondestructive evaluation (NDE) and pressure testing, the NC staff cannot find that the licensees proposed alternative provides for an adequate level of quality and safety. Therefore, the NC staff finds the licensees proposed alternative does not meet the requirements for U.S. Nuclear Regulatory Commission (NRC) authorization under Section 50.55a(z)(1) in Title 10 of the Code of Federal Regulations (10 CFR 50.55a(z)(1)), as requested.

Technical Evaluation The NC staff reviewed the licensees proposed alternate treatment process of the proposed alternative RA-22-0174 pursuant to 10 CFR 50.55a(z)(1). Specifically, the NC staff evaluated the licensees proposed alternative in this area to determine if it will provide an acceptable level of quality and safety.

Alternative Treatment In evaluating the licensees alternative treatment requirements of the proposed alternative, the NC staff considered the past precedent of previous NRC approved methods relating to risk-informed treatment of structures, systems, and components (SSCs) for nuclear power plants. As noted in the licensees submittal, these include previous NRC approval of the use of American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME) Code Case N-752 at Arkansas Nuclear One and 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. While the licensee has not requested to implement 10 CFR 50.69 at Oconee Nuclear Station (ONS) Units 1, 2 and 3, the licensee specified that the treatment requirements in its proposed alternative, which relies on ASME Code Case N-752, are consistent with scope of the requirements for similar Low Safety Significant (LSS) SSCs listed in 10 CFR 50.69(b)(1). While the NRC has not generically endorsed ASME Code Case N-752, this consistency for treatment to the NRC rules under 10 CFR 50.69, as specified by the licensee, was considered under NC staff review.

The NC staff also considered the operating experience of these programs in the review for the plant-specific application at ONS. The NC staff identified a concern related to the ASME Code Case N-752 application versus the 10 CFR 50.69 application, which include potential system versus individual item analysis, safety assessments differences and application differences. The NRC staff performed an independent risk assessment of this potential safety concern and found in accordance with Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, that the proposed alternative was acceptable for a risk-informed review of the alternative. However, the assessment also found the change in Core Damage Frequency values that were non-negligible in accordance with Figure 4 of RG 1.174. NC staff then considered the impact of defense-in-depth and safety margin, the second and third principles in RG 1.174, respectively.

II.2 Of note, in Section 2.1.2 of RG 1.174, it states, With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met Therefore, the NC staff performed a focused review on the potential impact of the licensees proposed alternative requirements relating to codes and standards for repair/replacement activities on safety margin.

The NC staff used the guidance of Federal Register Notice (FRN), 69 FR 68047, dated Nov. 22, 2004, for 10 CFR 50.69 that stated, In using consensus standards, the licensee or applicant must note that combining or omitting provisions of standards might result in ineffective implementation of 10 CFR 50.69 by causing RISC[Risk-Informed Safety Class]-3 SSCs to be incapable of performing their design basis safety functions. The NC staffs review of the specific language in the licensees proposed alternative identified several areas of uncertainty in providing reasonable assurance of meeting nationally approved codes and standards.

Reasonable Confidence Issue The NC staffs review of the specific use of the term reasonable confidence versus reasonable assurance identified an issue of uncertainty. The licensee listed their proposed treatment requirements for LSS items in Section 5.2.E of the submittal. ASME Code Case N-752, Paragraph -1420 requires the licensee to define alternative treatment requirements that confirm with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design-basis conditions. Neither ASME Code Case N-752 nor the licensees submittal specifically defines the term reasonable confidence. The licensee explained that this approach to treatment is consistent with RISC-3 treatment requirements specified in 10 CFR 50.69(d)(2). However, the 10 CFR 50.69 alternative treatments allow a licensee to exempt LSS items from 10 CFR Part 50, Appendix B, thereby using the term reasonable confidence in lieu of reasonable assurance as described in Section III.3.2 of the Statements of Consideration for the 10 CFR 50.69 rule (69 FR 68008). As neither ASME Code Case N-752 nor 10 CFR 50.55a(z)(1) allow a licensee to exempt LSS items from 10 CFR Part 50, Appendix B, the NC staff cannot find that the use of the term reasonable confidence rather than reasonable assurance establishes an acceptable level of quality and safety. The licensee uses the term reasonable confidence six times, Sections 5.1, 5.2.E, 5.2.E.12, 5.2.E.13, 5.2.E.16, and 5.2.F. The NC staff finds the use of reasonable confidence versus reasonable assurance for these sections raises uncertainty in evaluating the safety margin of the licensees proposed alternative treatment. As such, the NC staff cannot find that the licensees proposed alternative provides for an adequate level of quality and safety to ensure alternative treatments are adequate such that each LSS item remains capable of performing its safety-related functions under design-basis conditions.

Codes and Standards Alternative Issue The NC staffs review of the specific alternative codes and standards used also identified several areas of uncertainty in assessing the effective change in safety margin and thereby the quality of the proposed alternate treatment. The licensee listed the alternative treatments to meet Paragraph -1420 of ASME Code Case N-752 in Section 5.2.E. The NC staff found Sections 5.2.E.1 through 5.2.E.10 are equivalent to Subparagraphs -1420(a) through (j) of ASME Code Case N-752. The NC staff again notes that the NRC has not generically approved Code Case N-752, therefore the specific language of the licensees submittal was the focus of the NC staff review.

In reviewing the licensees specific alternative treatment wording of Section 5.2.E, the NC staff evaluated the alternative requirements in lieu of current regulatory requirements for codes and standards for the change in safety margin to evaluate the alternative treatment requirements for

II.4 in evaluating the change in safety margin of the alternative treatment that the staff cannot find that an acceptable level of quality and safety will be maintained.

Section 5.2.E.7 of the licensees submittal states the following, in part, Performance of repair/replacement activities, and associated NDE, shall be in accordance with the Owner's Requirements and, as applicable [emphasis added], 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. [emphasis added].

The NC staff finds that this language only requires the licensee to perform repair/replacement activities by the Owners Requirements with the option to follow the Construction Code or post-construction code or standard as deemed applicable by the Owner for the selected repair/replacement activity. Further, even if a code or standard is deemed applicable and chosen by the Owner for the selected repair/replacement activity, the allowance of alternative examination methods as approved by the Owner can significantly change the level of quality and safety relative to following a nationally approved code or standard. For example, a licensee may choose to substitute a volumetric examination with a visual examination that could be performed by a plant walkdown with insulation in place. This is a significant reduction in quality.

If a nationally approved code or standard for construction or post-construction clearly identifies an examination method for a repair/replacement activity, the option for the Owner to change that method raises significant uncertainty to the level of quality and change in safety margin for the performance of a repair/replacement activity.

Owners Requirements, as defined in the 2007 Edition with 2008 Addenda of Section XI of the ASME Code, are defined as those requirements when a construction code is not specified, address plant-specific requirements of the Construction Code, or invoke plant-specific requirements that are in excess of Construction Code requirements. However, the allowances of Sections 5.2.E.6 and 5.2.E.7 would allow the Owner to determine what codes and standards could be applied and then change specific provisions without adequate justification. The NC staff finds the allowance of the Owners options for performance of repair/replacement activities and NDE methods causes sufficient uncertainty in evaluating the change in safety margin of the alternative treatment that the staff cannot find that an acceptable level of quality and safety will be maintained.

In a supplemental audit response dated September 26, 2023, the licensee provided additional basis for the use of as applicable term in Section 5.2.E.7 by stating, Paragraph 5.2.E.7 states that performance of repair/replacement activities (reference IWA-4110b) will have both the Owners Requirements and either the Construction Code, or postconstruction code or standard, selected for the repair/replacement activity.

The NC staff note that reference IWA-4110b only describes the repair/replacement activity and does not use the term as applicable. While the licensees wording is one possible interpretation of the language of Section 5.2.E.7, it does not bound the possible options. The NC staff concern remains that only Owners Requirements are required for the performance of repair/replacement activities, and associated NDE.

In response to the wording Alternative examination methods may be used as approved by the Owner, the licensee states that, These paragraphs (the NDE portion of 5.2.E.7 and all of

II.5 5.2.E.8) permit specific alternatives to the code or standard, but not wholesale use of Owners Requirements to substitute for code requirements, and notes that these permissions are also subject to paragraph 5.2.E.3. The NC staff finds the additional information provided by the licensee does not provide sufficient technical information to justify the alternative to allow any alternative NDE method from a nationally approved code or standard when performing a repair/replacement activity. As highlighted in the Section 5.2.E.6 discussion above, ASME PCC-2, as a possible alternative post-construction code, does not allow a user to apply engineering judgement for NDE standards. However, the licensee states that this should be allowed due to wording in Section 5.2.E.3 which states, 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.

The NC staff finds that Section 5.2.E.7 is not clearly linked to the Section 5.2.E.3 changes requirement. As such, it is not specifically required to be performed or documented. The NC staff also observes that the wording of Section 5.2.E.3 would not require the item to maintain leak tightness, as it may seem to imply, because of the qualifying statement included for functional requirements of the system. Similarly, the NC staff considers that without a specific requirement, the Section 5.2.E.3 evaluation is insufficiently defined to provide reasonable assurance that a change in NDE method would ensure structural integrity for either a repair/replacement or mitigation for its design life. The NC staff finds the licensee does not provide a sufficient technical basis to allow the open-ended alternative of use of any NDE method, rather than those prescribed by a nationally approved code or standard for the particular repair/replacement method applied.

Section 5.2.E.8 of the licensees submittal states the following, 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 [emphasis added].

The NC staff notes the language would allow the Owner to choose what pressure testing of the repair/replacement activity will be performed rather than applying a nationally recognized construction, or post-construction code or standard. The NC staff finds this language to allow the Owner to choose that no pressure testing of the repair/replacement activity will be performed, even if it is a requirement of the Construction Code for the item. The NC staff finds the allowance of the Owners options for pressure testing of the repair/replacement activity causes uncertainty in evaluating the change in safety margin of the alternative treatment and that the staff cannot find that that alternative provides for an acceptable level of quality and safety.

In a supplemental audit response dated September 26, 2023, the licensee provided additional basis for the use of the language, or shall be established by the Owner, under Section 5.2.E.8.

This wording was included in the discussion of Section 5.2.E.7 above for the NDE method alternative but was also used to address the pressure testing alternative of Section 5.2.E.8.

Similar to the NC staff discussion above, the licensee did not provide a sufficient technical basis to allow the open-ended alternative of any option, including not performing any pressure test, regardless of the requirements of a nationally approved Construction Code for the particular repair/replacement method applied.

II.6 Consideration of Licensee Programs The NC staff, in light of the above identified concerns for an acceptable level of quality and safety, evaluated the audit responses from the licensee to an NRC request to clearly define the nationally approved codes and standards to be used as an alternative to the ASME Code requirements. In its responses, the licensee stated that the alternative requirements of ASME Code Case N-752 were more specific than those required by 10 CFR 50.69 for alternative treatment of LSS items. However, the NC staff notes that the process under 10 CFR 50.55a(z) to authorize the licensees proposed alternative to implement N-752, and the license amendment process to implement 10 CFR 50.69, are significantly different.

The NRC regulations in 10 CFR 50.55a(z) allow licensees to request alternatives to specific aspects of the ASME Code, which are typically much less extensive than the proposed request to implement ASME Code Case N-752 that would modify the treatment provisions for most Class 2 and 3 LSS items at a nuclear power plant. The Commission developed 10 CFR 50.69 to allow licensees to submit a license amendment request for detailed NRC staff review, which includes detailed probabilistic risk assessment methodology, system level evaluation process, specific nuclear SSCs safety classification requirements, and high-level treatment requirements for safety-related SSCs classified as LSS in lieu of certain special treatment requirements in the NRC regulations (such as 10 CFR Part 50, Appendix B). The specific implementation of processes specified in 10 CFR 50.69 and ASME Code Case N-752 is also very different. For example, the application of 10 CFR 50.69 to the full system for analysis of both active and passive functions of each component, Integrated Decision-making Panel review, and identification of non-ASME Code Class components for safety-related treatment is a more comprehensive process than the use of the licensees proposed alternative to utilize ASME Code Case N-752 for individual items. The NRC staff notes that this difference allows ASME Code Case N-752 to be applied quickly to address ongoing failures of components that require repair/replacement activities or to a system basis to apply non-ASME Code replacement or mitigation techniques.

Further, these applications of non-ASME Code repairs or mitigations can be performed throughout the majority of the ASME Code Class 2 and 3 boundaries for the period of the proposed alternatives authorized duration. The ASME Code technical basis document for the development of ASME Code Case N-752 Whitepaper in support of Item No. BC06-250, confirms the original option of ASME Code Case N-752 was to support an Owners implementation of 10 CFR 50.69 for the pressure boundary function and to be used independent of 10 CFR 50.69 for repair/replacement activities. Further the Whitepaper noted that in lieu of the current ASME Code requirements, a nationally recognized construction code or standard applicable to that item that is acceptable to the enforcement authority at the plant should be used. Changes through the ASME Code development process occurred until the final approval of ASME Code Case N-752. However, the NC staff finds that without establishing clear nationally recognized codes or standards to be used as an alternative to Section XI, the uncertainty to define the safety margin from IWA-4000 is too large. Thus, the NC staff cannot find that the licensees proposed alternative provides for an adequate level of quality and safety.

The NC staff evaluated the licensees proposed alternatives categorization process that uses a 1.0 failure probability to provide a conservative risk-informed categorization result to minimize uncertainty. While the NRC staff recognizes the conservative process to make the initial categorization of a component as High Safety Significant (HSS) or LSS, the NC staff used risk insights to evaluate the overall impact of the use of ASME Code Case N-752. The NC staff

II.7 notes that this conservative failure probability used in the ASME Code Case N-752 categorization is not used in future assessments of risk for that repaired or replaced component.

Instead, the failure probability is reset to a value associated with the history of the item under the previous regulatory and quality control requirements. Further, ASME Code Case N-752 can be used to address repair/replacement activities across multiple trains in the same safety system with no comprehensive system analysis to be performed. The NC staff believes the NRC must have reasonable assurance of the quality of the repair/replacement activity to justify utilizing this previous failure probabilities for future risk assessments of individual items within a safety system. The uncertainty raised by not requiring or defining the nationally approved codes and standards to be used in the repair/replacement methodologies and activities as identified in the NC staff review of the specific language of Section 5.2.E above, does not allow the NRC staff to find that an acceptable level of quality and safety will be maintained for the multiple uses of ASME Code Case N-752 as part of the licensees proposed alternative.

Additionally, the NC staff evaluated the licensee identified additional Owner responsibilities for other programs and processes that remain in place, such as design control, the 10 CFR 50.59 change control process, supply chain/procurement processes, corrective action/problem identification and resolution, testing and monitoring programs (e.g., risk-informed inservice inspection (RI-ISI), inservice testing, License Renewal Aging Management, Flow Accelerated Corrosion, Erosion, Raw Water Program, Buried Pipe Program, etc.), and Technical Specifications. While each of these programs provides additional defense-in-depth and monitoring, they will be impacted by the categorization of the applicable items under ASME Code Case N-752 as LSS items with undefined applicable construction and post-construction code allowances.

The NC staff notes that RI-ISI, for example, can have no required inspections of LSS items (including items categorized under Code Case N-752) unless an active degradation mechanism is identified, and then those inspection percentages are limited by the LSS status. The NC staff recognizes the effectiveness of the Technical Specification program, but also notes that ASME Code Case N-752 will allow non-Code repairs for defects in the safety systems for which the Technical Specifications are intended to ensure their operability. The NC staff finds that not requiring established nationally approved codes or standards reduces the defense-in-depth of the quality controls of these LSS categorized safety systems where repair/replacement activities are performed currently and for future defects in the same system. The NC staff notes that while no changes will occur to inservice testing or the maintenance rule due to licensee application of ASME Code Case N-752, no additional testing or maintenance will be required for active components that have had their pressure boundary items repaired without nationally approved codes or standards. Each of these programs has a risk consideration either in defense-in-depth or safety margin. The NC staff, utilizing risk insights to consider the impact of not requiring nationally approved codes or standards, has found the potential impact does not allow the staff to find that an adequate level of quality and safety will be maintained.

Conclusion Given the allowances provided by the licensee in Sections 5.2.E.6, 5.2.E.7 and 5.2.E.8 noted above for the use of national codes and standards for repair methods, repair/replacement activities, NDE and pressure testing, and the use of the term reasonable confidence in Sections 5.1, 5.2.E, 5.2.E.12, 5.2.E.13, 5.2.E.16, and 5.2.F without an exemption under 10 CFR 50.12 from 10 CFR Part 50 Appendix B, the NC staff finds the licensees proposed alternative does not allow the staff to determine that it will provide for an adequate level of quality and

II.8 safety. Therefore, the NC staff finds the licensees proposed alternative does not meet the requirements for NC authorization under 10 CFR 50.55a(z)(1) as requested.

The NC staff finds that a change to these sections identified above to clarify the implementation of the use of nationally approved codes and standards and the use of the term reasonable assurance will change these findings.

ENCLOSURE III.

BACKGROUND FOR OGC QUESTIONS 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Requirements Federal Register Notice (FRN), 69 FR 68047, dated November 22, 2004, (and SECY 04-0109, Final Rulemaking to Add New Section 10 CFR 50.69) provided a new regulation, Title 10 of the Code of Federal Regulations (10 CFR) 50.69, that provided an alternative approach for establishing the requirements for treatment of structures, systems, and components (SSCs) for nuclear power reactors using a risk-informed method of categorizing SSCs according to their safety significance. A licensee voluntarily choosing to implement this section shall submit an application for license amendment under 10 CFR 50.90 that contains the following information:

(i) A description of the process for categorization of RISC [Risk-Informed Safety Class]-1, RISC-2, RISC-3, and RISC-4 SSCs.

(ii) A description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific probabilistic risk assessment (PRA), margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

(iii) Results of the PRA review process conducted to meet § 50.69(c)(1)(i).

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy § 50.69(c)(1)(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions).

FRN 69 FR 68047,Section III.2.0, states that:

Before a licensee may implement § 50.69, the NRC must approve the categorization process through a license amendment. This is necessary because of the importance of the PRA and categorization process to successful implementation of the rule.

The Commission will approve a licensees implementation of this section if it determines that the process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs satisfies the requirements of 10 CFR 50.69(c) by issuing a license amendment approving the licensees use of this section. The FRN,Section III.2.0 states that the SSC categorization process must, Be performed for entire systems and structures, not for selected components within a system or structure. In addition,Section III.4.3 of the FRN states that, The Commission will not remove the repair and replacement provisions of the ASME [American Society of Mechanical Engineers]

Code required by 10 CFR 50.55a(g) for ASME Class 1 SSCs, even if they are categorized as RISC-3, because those SSCs constitute principal fission product barriers as part of the reactor coolant system or containment.

The Commission specified the implementation of 10 CFR 50.69 for RISC component classification in the FRN on page 68033 as follows:

III.2 The pilot experiences also revealed the intricacies of the relationship between functions (which play a role in decisions on safety significance) and components (importance measures are associated with components and treatment is also generally applied on a component basis). Because a particular component may support more than one function, the categorization of the component needs to correspond with the most significant function and means must be provided for a licensee to map the components to the functions they support.

In support of its position at a recent public meeting, the industry referenced the safety evaluation on the Vogtle Units 1 and 2 license amendment request (LAR) for the use of 10 CFR 50.69. On electronic page 16, the safety evaluation references the licensee response to Request for Additional Information (RAI) #29. On electronic page 16 of its response to RAI #29, Southern Nuclear Company (SNC) states the following:

The NEI 00-04 categorization methodology assigns risk at the component level. Per the methodology, a component gets assigned final risk if any of the following risks is HSS: active risk, passive risk, or defense in depth.

10 CFR 50.69s passive categorization method is defined, in part, by NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, which was endorsed in Regulatory Guide (RG) 1.201, Guidelines For Categorizing Structures, Systems, And Components In Nuclear Power Plants According To Their Safety Significance, which is For Trial Use and includes Section 10.2, which states:

The necessity of addressing each component or each part of a component is determined by each licensee based on the anticipated benefit. A licensee may determine that it is sufficient only to perform system or subsystem analyses, RISC categorizing all SSCs within a system or subsystem according to whether the system or subsystem as a whole performs a risk significant function (Section 10.1). In such cases, all the components within the boundaries of the subsystem or system would be governed by the same set of safety-significant functions.

Each licensee has the option, based on the estimated benefit, of performing additional engineering and system analyses to identify specific component level or piece part functions and importance for the safety-significant SSCs.

Industrys position is that this allows categorization of a High Safety Significant (HSS) active component by separating out each piece of the component (i.e., valve bonnet, valve stem, etc.)

and categorizing some piece parts as Low Safety Significant (LSS) passive components. In other words, a valve bonnet can be classified as LSS passive while the valve stem is classified as HSS active. However, the staff notes the following:

NRC interpretation in comments in Draft Guide (DG)-1121, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (ML031430373) to RG 1.201 concerning Section 10.2 (NEI-00-04) was that For licensees that perform the optional step, Detailed Engineering Review of HSS Components, the same depth and rigor must be used in categorizing at the individual component level as was used for categorizing at the system functional level. Further NRC interpretation in DG-1121 was that The specific considerations that permit a LSS determination of an SSC in a

III.4 10 CFR 50.69 is the regulation that allows removal of all regulatory requirements for repair/replacement and quality control for LSS SSCs.

NRC staff previously approved an alternative by safety evaluation in 2009 (ML090930246) to use ASME Code Case N-752, Revision 0, for Arkansas Nuclear One, Units 1 and 2 (ANO), but with additional Quality Assurance (QA) requirements. The NRC staff also approved another alternative by safety evaluation, dated May 19, 2021 (ADAMS Accession No. ML21118B039) for using ASME Code Case N-752 for the remaining license renewal period. The NRC staff reviewed the plant-specific PRA in granting this alternative. In addition, the NRC staff approved the licensee changes to the Quality Assurance Procedure Manual (QAPM) that SSCs categorized as LSS will not be required to meet the requirements of the QAPM. Instead, Entergy would develop program elements describing treatment of these LSS SCCs to ensure continued capability and reliability of the design-basis function. There are several other alternatives to use ASME Code Case N-752 currently under review by the NRC staff. Note that the alternatives stated, The categorization and treatment requirements of Code Case N-752 are consistent with those in 10 CFR 50.69.

Neither ASME Code Case N-752 or N-752-1 have been included for use in RG 1.147. ASME Code proposed changes to ASME Code Case N-752-1, which were incorporated into N-752-2, based on the pilot ANO relief request, lessons learned from the pilot, and related industry efforts.

In reviewing Code Case N-752-1 for including in the current rulemaking for 10 CFR 50.55a, the Office of the General Council (OGC) had the following concerns:

1. Cannot mandate the use of RG 1.200, Revision 3 without incorporating by reference the RG into the rule. Currently, RG 1.200 is not ready for incorporating by reference in the RG.

2. If we approve this Code Case and include it in RG 1.147, which will be incorporated by reference in 10 CFR 50.55a, does the use of the Code Case by licensee deviate from the Commission direction that a license amendment is needed to apply risk-informed categorization per 10 CFR 50.69? Once incorporated by reference in 10 CFR 50.55a, a licensee can apply N-752-1 without NRC approval. A related issue is:

a. If we do not approve this Code Case, licensee will use this through the 50.55a alternative process, 50.55a(z)(1). If we approve these alternative requests, does the approval deviate from the Commission direction that a license amendment is needed to apply risk-informed categorization per 10 CFR 50.69?

The staff notes that ASME Code Cases within 10 CFR 50.55a do not allow a licensee to be exempt from NRC special treatment requirements in other portions of the NRC regulations, such as the quality assurance requirements in 10 CFR Part 50, Appendix B. In addition, 10 CFR 50.54 states:

The following paragraphs of this section, with the exception of paragraphs (r) and (gg), and the applicable requirements of 10 CFR 50.55a, are conditions in every nuclear power reactor operating license issued under this part.

10 CFR 50.54(a)(3)(ii) allows a quality assurance alternative or exception approved by an NRC safety evaluation, provided that the bases of the NRC approval are applicable to the licensees facility by including it in an updated final safety analysis report.

III.6 Section 10.2 Detailed SSC Categorization The necessity of addressing each component or each part of a component is determined by each licensee based on the anticipated benefit. A licensee may determine that it is sufficient only to perform system or subsystem analyses, RISC categorizing all SSCs within a system or subsystem according to whether the system or subsystem as a whole performs a risk significant function (Section 10.1). In such cases, all the components within the boundaries of the subsystem or system would be governed by the same set of safety-significant functions. Each licensee has the option, based on the estimated benefit, of performing additional engineering and system analyses to identify specific component level or piece part functions and importance for the safety-significant SSCs.

Neither ASME Code,Section XI nor ASME Code Case N-752 specify the term piece-part. During a recent public meeting, industry representatives stated that the term SSC includes piece-part, and that it includes subcomponent parts of a component.

The staff position is that the term piece-part was for the further evaluation for categorizing a component in a system, not piece parts of a component, but rather piece parts of system. This is further explained in NRC interpretation to comments in Draft Guide (DG)-1121 (ML031430373) of RG 1.201 of Section 10.2 (NEI-00-04) which stated that For licensees that perform the optional step, Detailed Engineering Review of HSS Components, the same depth and rigor must be used in categorizing at the individual component level as was used for categorizing at the system functional level. Further interpretation was that The specific considerations that permit a LSS determination of an SSC in a safety-significant functional flow path must not be limited to just active failure modes but must consider all potential failure modes for the subject SSC. In addition, 10 CFR 50.69 Federal Register Notice (69 FR 68008, 68033) provides the Commissions intent that 10 CFR 50.69 RISC classification would be performed at the component level.

Technical Concern The staff has concerns on how licensees categorize the overall safety significance of components applying ASME Code Case N-752 to the passive pressure retaining function of a component without considering the impact of the relaxed treatment of the components relating to the passive function and how it may interact with the retained level of treatment of the active function.

For the case where a single component can have both an LSS categorization of passive function and a HSS categorization of its active function, the licensee stated that only the passive function will get relaxed treatment. A staff concern is the impact of the relaxed treatment of the subcomponents relating to the passive function and how it may interact with the retained level of treatment of the active function. The concerns include:

III.7 A component relies on each part of the component to perform its function. A passive part can affect the active part of a component. For example:

o Valve bonnet (passive) can affect the valve stem (active) and prevent or slow the closing or opening of the valve.

o Valve seat (passive) can affect the valve disc (active) from properly seating or damage the valve seat or disc, or proper opening of the valve.

o Valve body (passive) could fail and divert the intended flow capability or intended flow isolation in a manner to preventing function (active) of the valve.

Nuclear plant operating experience has shown that the valve bonnet can interfere with the motion of the valve stem in motor-operated valves (MOVs).

If the valve bonnet is classified as an LSS piece-part, the alternative repair and replacement activities performed by the licensee might not have sufficient controls to ensure that the active function of the valve is not impacted because the active and passive parts interact and connect to each other.

Applying separate treatment requirements for each subpart of a component would be extremely difficult to control safely. For example, an MOV actuator has over 100 subparts. It would be difficult to determine whether the passive function of each of those subparts might have an impact on the active function of the actuator.

In order to determine whether the alternative treatment provides reasonable confidence that the component will perform its function, the specific alternate treatment needs to be specified and used in its entirety and shall be applicable to the specific component and must be taken into consideration in the categorization process. ASME Code Case N-752 only states that nationally recognized codes and standards (e.g., ASME PCC-2, Repair of Pressure Equipment and Piping, API-653, Tank Inspection, Repair, Alteration and Reconstruction) may be used or as approved by the Owner, which would allow any requirement. Note that API-653 specified for use in ASME Code Case N-752 is for carbon steel storage tanks but could be used for a stainless-steel valve per ASME Code Case N-752. ASME B31.1, Power Piping, is more applicable to piping, valves, etc. that would encompass ASME Code Case N-752. As such, ASME Code B31.1 and ASME Code PCC-2 are codes that the Staff has reasonable confidence will provide adequate treatment, and a more defined change in safety margin, of these categorized LSS components since they were used in balance of plant systems for nuclear sites. It should be noted that FRN 69 FR 68047,Section V.5.2 provides the following:

However, as described in NUREG/CR-6752, A Comparative Analysis of Special Treatment Requirements for Systems, Structures, and Components (SSCs) of Nuclear Power Plants with Commercial Requirements of Non-Nuclear Power Plants, significant variation exists in the application of industrial practices at nuclear power plants. Hence, a simple reference to these practices does not provide a basis to satisfy the rules requirements. To satisfy the requirement that the treatment of RISC-3 SSCs be consistent with the categorization process, the licensee or applicant must establish treatment processes that provide reasonable confidence that SSCs perform their safety-related functions consistent with reliability levels used in the categorization process. The licensee or applicant must either establish treatment processes that provide this level of reliability or use consensus standards that provide a proven level of reliability based on experience. In using consensus standards, the licensee or applicant must note that combining or omitting provisions of standards might result in ineffective implementation of § 50.69 by causing

III.8 RISC-3 SSCs to be incapable of performing their design basis safety functions. The NRC considers the ASME code cases endorsed in § 50.55a and listed in RG 1.84, 1.147, and 1.192 to be one acceptable method of establishing treatment of RISC-3 SSCs, where applicable, in that those applicable endorsed code cases adjust treatment based on the safety significance of the components.

This provides the basis that the licensee must establish alternative treatment in order to provide reasonable confidence that SSCs perform their safety-related functions.

Based on the above, the NC staff request OGC assistance in resolving the following questions related to the request by Duke Energy under 10 CFR 50.55a(z)(1) to use ASME Code Case N-752 at Oconee Nuclear Station (ONS):

1. Confirm that the Staff cannot authorize the use of ASME Code Case N-752 (which the licensee states is consistent with 10 CFR 50.69) under the 10 CFR 50.55a(z) alternative process because it bypasses 10 CFR 50.69(b)(2) in which a license amendment is required. In other words, confirm that recategorization of SSCs to high-safety significant (HSS)/LSS and subsequent changes in special treatment of safety-related SSCs is only permitted under the auspice of 10 CFR 50.69.

2. ASME Code Cases within 10 CFR 50.55a do not allow a licensee to be exempt from NRC special treatment requirements in other portions of the NRC regulations, such as the QA requirements in 10 CFR Part 50, Appendix B. Confirm that 10 CFR 50.55a and 10 CFR 50.54 does not allow a licensee to be exempt from the quality assurance requirements in 10 CFR Part 50, Appendix B.

3. Confirm that the previous plant-specific authorization under 10 CFR 50.55a(z) to use ASME Code Case N-752 for ANO does not constitute a past precedent that would require the evaluation of backfit issues for the use of ASME Code Case N-752 at ONS.

Summary of Issues (SOI) - NCP-2023-005 Issue 1 Duke Energy has requested authorization of their proposed alternative RR-22-0174 to use American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME) Code Case N-752, Risk Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1, at Oconee Nuclear Station (ONS), Units 1, 2, and 3, without exception or deviation. Under these conditions, the NC staff's understanding is for Duke Energy to implement their proposed alternative of Section 5.2.E.14, As permitted by Code Case N-752, Duke Energy intends to implement the QA [quality assurance] Program exemption applicable to IWA-1400(n), at ONS, they would need an exemption from 10 CFR Part 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, for those applicable low safety significant (LSS) items. Additionally, Duke Energy uses the term reasonable confidence to describe the level of the capability of safety-related items to meet their safety function. Neither the licensee nor ASME Code Case N-752 defines reasonable confidence. Based on the development of ASME Code Case N-752, the NC staff understands that the term reasonable confidence is only explained in the Federal Register Notice for 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors. Under 10 CFR 50.69, for a licensee to apply the standard of reasonable confidence, a licensee is expressly exempt from the requirements of 10 CFR Part 50, Appendix B, only upon issuance of the license amendment to apply 10 CFR 50.69 at a plant. As a further complication, the NC staff believe that Entergy Operations, Inc.

(Entergy) has assumed that they received just such an exemption based on its previous authorization to use ASME Code Case N-752 and the QA manual change at Arkansas Nuclear One, Units 1 and 2 (ANO). The NC staff believe this assumption by Entergy is made clear by statements in several licensee documents as shown in Enclosure I to Section A the NCP-2023-005. Through the audit process, the NC staff are concerned that Duke Energy will utilize 10 CFR 50.54, Conditions of licenses, paragraph (a)(3)(ii), as a basis, to claim this same perceived exemption as Entergy believes they have obtained. The NC staff notes that neither Entergy nor Duke Energy has, or is, requesting explicit exemption from Appendix B under 10 CFR 50.12, Specific exemptions, for those applicable ASME Code Case N-752 items at ANO or ONS, respectively. The NC staff believe such an exemption is necessary prior to implementing the proposed alternative or Duke Energy should modify their proposed alternative. If neither option is employed, the NC staff are concerned that regulatory uncertainty would result from the inconsistency between the understanding of the applicability of 10 CFR Part 50, Appendix B, by the NRC staff and licensees. For details and specific statements by Entergy and Duke Energy that have led to this safety concern, see Enclosure I to Section A of NCP-2023-005.

Issue 2 The NC staff identified a concern related to the quality of the alternative treatment options in the Duke Energy proposal in lieu of the current regulatory requirements for repair/replacement activities. To address this concern, the NC staffs review of the specific language in the licensees proposed alternative identified several areas of uncertainty in providing reasonable assurance of meeting nationally approved codes and standards. This was the basis for audit and request for additional information (RAI) questions to Duke Energy requesting the licensee to identify the specific nationally approved codes and standards for their proposed alternative. The NC staff found Duke Energys response to not provide specific codes and standards led to Section C: Summary of Issues

further uncertainty in assessing the safety margins and effectiveness of defense-in-depth measures. Therefore, the NC staff cannot find that the licensees proposed alternative provides for an adequate level of quality and safety. In Enclosure II to Section A of NCP-2023-005, the NC staff provide their basis for denial of the licensees proposed alternative request. The NC staff continue to believe that this issue could be addressed if the licensee established minimum requirements for specific codes and standards for repair/replacement activities associated with the implementation of the plant specific proposed alternative at ONS.

Issue 3 For both of the above issues, the NC staff believe that OGC review of the safety evaluation with consideration of specific questions, listed below, would help address the difference in opinions between the NC staff and other NRC staff and management on these issues.

As allowed by Section 5 in LIC-102, Revision 3, Review of Relief Requests, Proposed Alternatives, and Requests to Use Later Code Editions and Addenda, the technical branch chief and subject matter experts can request OGC legal review of relief/alternative requests on the basis of perceived unique or special circumstances. This is a unique review with special circumstances that was initially brought to the attention of the NRC staff by OGC during the 10 CFR 50.55a, Codes and Standards, rulemaking for incorporating ASME Code Cases, including ASME Code Case N-752, into 10 CFR 50.55a. The OGC comment on ASME Code Case N-752 included whether the use of the Code Case by a licensee deviates from the Commission direction that a license amendment is needed to apply risk-informed categorization per 10 CFR 50.69. This is due to the fact that the technical basis for the licensees safety categorization process to utilize ASME Code Case N-752 is based on it being the same as the passive categorization method of 10 CFR 50.69. Note that the proposed alternative states, The categorization and treatment requirements of Code Case N-752 are consistent with those in 10 CFR 50.69. The NRC regulations in 10 CFR 50.69 allow removal of certain special treatment requirements, such as 10 CFR Part 50, Appendix B for repair/replacement and quality control for safety-related LSS structures, systems, and components (SSCs). This also includes the policy issue of whether approving alternative requests for use of ASME Code Case N-752 deviates from the Commission direction that a license amendment is needed to apply risk-informed categorization per 10 CFR 50.69. Therefore, the DORL safety evaluation should be reviewed by OGC to determine if this contradicts the Commission's policy, which would also include OGC assistance in resolving the following questions related to the request by Duke Energy under 10 CFR 50.55a(z)(1) to use ASME Code Case N-752 at ONS:

1. Confirm that the Staff cannot authorize the use of ASME Code Case N-752 (which the licensee states is consistent with 10 CFR 50.69) under the 10 CFR 50.55a(z) alternative process because it bypasses 10 CFR 50.69(b)(2) in which a license amendment is required. In other words, confirm that recategorization of SSCs to high-safety significant (HSS)/LSS and subsequent changes in special treatment of safety-related SSCs is only permitted under the auspice of 10 CFR 50.69.

2. ASME Code Cases within 10 CFR 50.55a do not allow a licensee to be exempt from NRC special treatment requirements in other portions of the NRC regulations, such as the QA requirements in 10 CFR Part 50, Appendix B. Confirm that 10 CFR 50.55a and 10 CFR 50.54 does not allow a licensee to be exempt from the QA requirements in 10 CFR Part 50, Appendix B.

3. Confirm that the previous plant-specific authorization under 10 CFR 50.55a(z) to use ASME Code Case N-752 for ANO does not constitute a past precedent that would require the evaluation of backfit issues for the use of ASME Code Case N-752 at ONS.

For additional basis and background on these questions, please see Enclosure III to Section A of NCP-2023-005.

Issue 1 The Safety Evaluation for the use of ASME Code Case N-752 at Oconee Nuclear Station (ONS) does not grant an exemption from NRC regulations.

As noted in Code Case N-752 and acknowledged in the non-concurrence, while the code case states that low safety significance items do not have to meet the requirements of IWS-1400(o),

Footnote 1 states If compliance with 10 CFR 50 Appendix B or NQA-1 is required at the owners facility, IWA-1400(o) is not exempt.

ONS did not request an exemption from appendix B and has a quality assurance program description (QAPD) that meets appendix B. However, the requirements for implementation of and changes to quality assurance plans subject to appendix B of 10 CFR Part 50 are held in 10 CFR 50.54(a).

50.54(a)(1) requires, in part, that Each nuclear power plant...subject to the quality assurance criteria in Appendix B of this part shall implementthe quality assurance program described or referenced in the safety analysis report, including changes to that report.

50.54(a)(3) requires, in part, that Each licensee described in paragraph (a)(1) of this section may make a change to a previously accepted quality assurance program description included or referenced in the Safety Analysis Report without prior NRC approval, provided the change does not reduce the commitments in the program description as accepted by the NRCIn addition to quality assurance program changes involving administrative improvements and clarifications, spelling corrections, punctuation, or editorial items, the following changes are not considered to be reductions in commitment:

50.54(a)(3)(ii) describes that The use of a quality assurance alternative or exception approved by an NRC safety evaluation, provided that the bases of the NRC approval are applicable to the licensee's facility; is not considered to be a reduction in commitment.

50.54(a)(4) requires that Changes to the quality assurance program description that do reduce the commitments must be submitted to the NRC and receive NRC approval prior to implementation.

On May 19, 2021, Arkansas Nuclear One (ANO) was approved for plant-specific use of the ASME Code Case N-752 requirements. Concurrent with their request for the use of CC N-752, ANO submitted a quality assurance program manual change for NRC review and approval per 50.54(a)(4). This submittal was a reduction in commitments because it requested that Class 2 and 3 SSCs identified as LSS in accordance with N-752 not be required to meet the QAPM.

The submittal contained program elements describing the treatment of LSS SSCs to ensure continued capability and reliability of the design-basis function. In its approval, the NRC staff concluded that there is reasonable assurance that the licensees implementation of ASME Code Case N-752 will ensure that Class 2 and 3 LSS SSCs will perform their intended safety-related functions under design basis conditions and that the proposed Entergy QAPM change continues to provide an acceptable level of quality and safety.

Section C: Evaluation of Non-Concurrence

In their submittal for use of CC N-752, Duke Energy states that they will update their fleet QAPD for safety-related Class 2 and 3 SSCs identified as LSS in accordance with the code case to not be required to meet the QAPD and that they will develop elements describing treatment of these LSS SSCs to ensure continued capability and reliability of the design basis function. These requested changes are consistent with the approved SE from ANO, which is explicitly allowed by 50.54(a)(3)(ii). Duke does not need to request an exemption to 10 CFR Appendix B under 10 CFR 50.12 because the regulations in 50.54(a) explicitly describe the process for licensees to seek review and approval for reductions in commitments in their QAPDs (as ANO did) or for other licensees to use a quality assurance alternative or exception approved by an NRC safety evaluation, provided that the bases of the NRC approval are applicable to the licensee's facility (as ONS did).

With regard to the use of the term reasonable confidence, the non-concurring staff are concerned that this application of this term, used in 50.69, indicates an assumed exemption from the requirements of 10 CFR 50 Appendix B. 69 FRN 68008 Section III.3.2 describes reasonable confidence in the context of RISC-3 treatment in 50.69 as a somewhat reduced level of confidence as compared with the relatively high level of confidence provided by the current special treatment requirements. This flexibility is in recognition of the lower safety significance of the SSCs. The assessment of whether a LSS SSC would perform its safety-related functions under design basis conditions is not dependent on the term reasonable assurance vs. reasonable confidence, but rather in the controls that the licensee will put in place for the Class 2 and Class 3 components designated as LSS per N-752.

In its SE input on Quality Assurance, the NRC staff found that Duke plans to use its QAPD processes and procedures with controls for Class 2 and 3 LSS SSCs to ensure continued capability and reliability of the design-basis function. In addition, the changes proposed by Duke are consistent with the changes approved by the NRC staff to Entergys QAPD, which concluded that there is reasonable assurance that the licensees implementation of ASME Code Case N-752 will ensure that Class 2 and 3 LSS SSCs will perform their intended safety-related functions under design basis conditions and that the proposed Entergy QAPM change continues to provide an acceptable level of quality and safety.

Issue 2 In determining the need for a licensee to specify specific codes and standards as requested by the non-concurring employees vs. using nationally recognized codes and standards, as described in N-752, several areas were considered.

The Commission directed the staff to apply risk-informed principles in any licensing review or other regulatory decision in SRM-SECY-19-0036-Application of the Single Failure Criterion to NuScale Power LLCs Inadvertent Actuation Block Valves.

In any licensing review or other regulatory decision, the staff should apply risk-informed principles when strict, prescriptive application of deterministic criteria such as the single failure criterion is unnecessary to provide for reasonable assurance of adequate protection of public health and safety.

The intent of ONS's request to use N-752 is to allow for an alternative to the ASME Code requirement by allowing for more flexibility for the procurement and treatment of components

identified as LSS. The alternatives approach is to apply, in place of Section XI of the ASME Code, the requirements from the original Construction Code, Owners Requirements, and nationally recognized codes, standards, or specifications applicable to the LSS categorized item as permitted by the licensing basis.

The non-concurring staff's concern is that allowance for the use of Owners options in lieu of specific codes and standards, as proposed, causes "significant uncertainty in evaluating the safety margin and a concern to the quality of the licensees proposed alternative treatment." In essence, the non-concurring staff believes that such latitude could result in the licensee inappropriately applying standards that the NRC may/would not permit as alternatives, although neither the NRC nor licensee has identified a realistic scenario where this would occur at ONS.

Further, the review of ONS's submittal is not for a component-specific relief but rather for a general methodology/approach to using alternative treatment for LSS components.

As such, in applying the risk-informed principles laid out by the Commission in SRM-19-0036, and absent a clear safety concern, I find it unnecessary to be so prescriptive in identifying a specific standard to arrive at a reasonable assurance finding of adequate protection of public health and safety.

Further, it is the licensees responsibility to ensure that components perform their design and safety functions. The licensee describes in their request that they will define treatment requirements to address design control, procurement, installation, configuration control and corrective action. We have confidence that these controls, in combination with the low-risk significance of the components in question, will result in adequate safety.

Issue 3 With respect to the legal questions raised, it is acknowledged that there are provisions in LIC-102 for OGC legal review of relief/alternative requests.

Proposed Question 1:

We do not intend to bring this question to OGC for review.

The proposed question to OGC assumes that the staff cannot approve the use of ASME Code Case N-752 under the 10 CFR 50.55a(z) alternative process because it bypasses 10 CFR 50.69(b)(2) where a license amendment is required. However, the NCP did not make a case to show where either 50.69 or 50.55a explicitly or implicitly prohibits the NRC from doing so.

Rather, the NCP likens the ONS request to a 50.69 request and thus insists that it must be done via a LAR.

The intent of 10 CFR 50.69 is to provide a voluntary alternative framework for licensees to risk-inform their categorizing of SSCs according to their safety significance. This shouldnt be misunderstood to mean that any and all risk-informed approaches be done via a 50.69 LAR submittal. In ONSs case, the licensee is proposing to use an alternative to the ASME Code requirements for treatment of LSS passive components, including the use of aspects of methodology defined under 50.69 to arrive at the HSS/LSS categorization. While the licensee is leveraging the methodology used to meet 50.69, through NEI-00-04, to perform the categorization, the staff should not overextend that adoption of methodology to be equivalent to a request under 50.69.

Proposed Question 2:

We do not intend to bring this question to OGC for review. Approval of an alternative/relief within 50.55a does not constitute an exemption to other NRC regulations. Neither the licensee nor any NRC staff are proposing that this would be an exemption to 10 CFR 50 Appendix B.

50.54(a), as discussed in detail in our response to Issue 1, sets the requirements for implementing and documenting QA plans, as well as making changes to QA plans.

Proposed Question 3:

We do not intend to bring this question to OGC for review but rather consulted with the backfit community of practice on this question. The past approval of Code Case N-752 for ANO would not require a backfit analysis for a decision on ONS. However, deviation from precedent and change in Agency position should be approached cautiously as the staff should have a clear safety basis for arriving at a different decision for similar requests.