ML25160A235

DPO Case File: DPO-2024-002--Redacted-Public

ML25160A235
Person / Time
Site:	Oconee
Issue date:	06/09/2025
From:	Figueroa-Toledo G, Jordan Hoellman, Mark King Office of Nuclear Reactor Regulation
To:	Jay Collins, Stephen Cumblidge, John Honcharik, Matthew Mitchell, Thomas Scarbrough, Dan Widrevitz Office of Nuclear Reactor Regulation
References
DPO-2024-002
Download: ML25160A235 (0)

Text

DPO Case File for DPO-2024-002 The following pdf represents a collection of documents associated with the submittal and disposition of a differing professional opinion (DPO) from NRC employees regarding the NRCs authorization of Duke Energys alternative request to use ASME Code Case N-752 at Oconee Nuclear Station, Units 1, 2, and 3.

Management Directive (MD) 10.159, NRC Differing Professional Opinions Program, describes the DPO Program. https://www.nrc.gov/docs/ML2312/ML23123A099.pdf The DPO Program is a formal process that allows employees and NRC contractors to have their differing views on established, mission-related issues considered by the highest-level managers in their organizations, i.e., Office Directors and Regional Administrators. The process also provides managers with an independent, three-person review of the issue (one person chosen by the employee).

Because the disposition of a DPO represents a multi-step process, readers should view the records as a collection. In other words, reading a document in isolation will not provide the correct context for how this issue was reviewed and considered by the NRC.

It is important to note that the DPO submittal includes the personal opinions, views, and concerns of several NRC employees. The NRCs evaluation of the concerns and the NRCs final position are included in the DPO Decision.

The records in this collection have been reviewed and approved for public dissemination.

Document 1: DPO Submittal Document 2: Memo Establishing DPO Panel Document 3: DPO Panel Report Document 4: DPO Decision

Document 1: DPO Submittal

Page 1 of 1 NRC FORM 680 (09-2019)

NRC FORM 680 U.S. NUCLEAR REGULATORY COMMISSION (09-2019)

NRC MD 10.159 DIFFERING PROFESSIONAL OPINION DPO Case Number DPO-2024-002 Date Received 05/07/2024 Name(s) and Title(s) of Submitter(s)

Matthew Mitchell - BRANCH CHIEF; John Honcharik -

SENIOR MATERIALS ENGINEER; Jay Collins - SENIOR MATERIALS ENGINEER; Thomas Scarbrough - SENIOR MECHANICAL ENGINEER; Dan Widrevitz - MATERIALS ENGINEER; Stephen Cumblidge - MATERIALS ENGINEER Organization NRR Work Email Matthew.Mitchell@nrc.gov; John.Honcharik@nrc.gov; Jay.Collins@nrc.gov; Thomas.Scarbrough@nrc.gov; Dan.Widrevitz@nrc.gov; Stephen.Cumblidge@nrc.gov Name and Title of Supervisor Matthew Mitchell - BRANCH CHIEF; Samuel Lee - ; Angie Buford - ; Stewart Bailey - BRANCH CHIEF; Brian Smith -

NRR - DIRECTOR, DIV NEW AND RENEWED LIC.

Organization NRR Work Email Matthew.Mitchell@nrc.gov; Samuel.Lee@nrc.gov; Angela.Buford@nrc.gov; Stewart.Bailey@nrc.gov; Brian.Smith@nrc.gov When was the prevailing staff view, existing decision, or stated position established?

12/13/2023 Where (i.e., ADAMS ML#, if applicable):

Accession No. ML23262A967 Subject of DPO NRC Authorization Of Duke Energy Alternative Request To Use Of ASME Code Case N-752 At Oconee Nuclear Station, Units 1, 2, And 3 Summary of prevailing staff view, existing decision, or stated position; Reason for DPO, potential impact on mission, and proposed alternatives.

See Attachment Describe the (a) importance of prompt action on the issue, (b) safety significance of the issue, and (c) the complexity of the issue.

Do you believe the issue represents an immediate public health and safety concern?

No Is the issue directly relevant to a decision pending before the Commission?

No If Yes, Reference Document (i.e., ADAMS ML#):

Did informal discussions take place?

Yes If Yes, with whom and during what time frame?

See Non-concurrence (last quarter 2023) NCP-2023-005 Proposed panel members are:

Sheldon Clark (He/Him); Robert Tregoning; Kevin Coyne List of area(s) of technical expertise needed to properly assess the issue (e.g., electrical engineering, operator licensing).

OGC Materials engineering When the process is complete, I would like management to determine whether public release of the DPO case file (with or without redactions) is appropriate (Select No if you would like the DPO case file to be non-public):

Yes Please note that your DPO submittal may be shared on a need-to-know basis in an effort to resolve the concern, determine the most appropriate regulatory actions in response to the concern, and identify key agency resources to evaluate the concern.

DPO Accepted:

DPO Accepted / Rejected By DPO Accepted / Rejected On 05/07/2024 Gladys Figueroa-Toledo

DIFFERING PROFESSIONAL OPINION NRC AUTHORIZATION OF DUKE ENERGY ALTERNATIVE REQUEST TO USE OF ASME CODE CASE N-752 AT OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 COMPLETED LICENSING ACTIVITIES NRC safety evaluation dated December 13, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23262A967), authorized Duke Energy Carolinas, LLC (Duke Energy) alternative request for Oconee Nuclear Station (ONS), Units 1, 2, and 3 and the Keowee Hydro Station, Units 1 and 2, provided in its letter dated July 27, 2022 (ML22208A031), as supplemented by letters dated March 9, 2023 (ML23068A015), and October 20, 2023 (ML23293A267). Specifically, Duke Energy requested to use the alternative requirements of 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, for determining the risk-informed categorization and for implementing alternative treatment requirements for repair/replacement activities on moderate and high energy Class 2 and 3 items in lieu of ASME Boiler and Pressure Vessel Code (BPV Code),Section XI, paragraph IWA-1000, IWA-4000, and IWA-6000 requirements. This alternative request was submitted and authorized under Title 10 of the Code of Federal Regulations (10 CFR) in Section 50.55a, Codes and Standards, paragraph (z)(1) as providing an acceptable level of quality and safety.

The NRC staff authorization of the ONS alternative request was based on Entergy Operations, Inc. (Entergy) precedent for Arkansas Nuclear One (ANO) to use Code Case N-752. On May 19, 2021, the NRC (ML21118B039) authorized Entergy alternative request EN-20-RR-001, dated May 27, 2020 (ML20148M343), as supplemented by letter dated November 17, 2020, to use Code Case N-752. As part of this alternative, Entergy submitted a request to reduce its commitments in the Quality Assurance Program Manual (QAPM) at ANO during implementation of Code Case N-752 under 10 CFR 50.54, Conditions of licenses, on October 26, 2020, as supplemented by letters dated April 5, 2021, and April 30, 2021. On May 19, 2021 (ML21132A279), the NRC staff issued an SE approving the proposed change to the QAPM at ANO under 10 CFR 50.54 with specific quality assurance controls (although less rigorous than previously applied) for safety-related Class 2 and 3 components categorized as low safety significant (LSS) when implementing Code Case N-752.

NRC staff submitted non-concurrence (NC) NCP-2023-005, Non-concurrence of SE for Proposed Alternative RR-22-0174 for Oconee Nuclear Station, Units 1, 2 and 3 Duke Energy Carolinas, LLC (ML23262A967). The purpose of this Differing Professional Opinion (DPO) is to clarify, with additional details, the basis of the NC and to elevate concerns (in Enclosure I) with the DORL management response to the NC for further NRC Management review through the NRC DPO Process.

INTRODUCTION ASME Code Case N-752 would allow a licensee to determine the risk-informed categorization of Class 2 and 3 items and to use alternative treatment requirements for repair/replacement activities for Class 2 and Class 3 items determined to be LSS in lieu of repair/replacement requirements in ASME Boiler and Pressure Vessel Code (BPV Code),Section XI and without

2 applying the quality assurance (QA) requirements of 10 CFR Part 50, Appendix B. When the phrase repair/replacement requirements is used, it would include materials, fabrication, use of construction code (i.e.,Section III, etc.) for design and construction, non-destructive inspection and pressure testing. Industry presentations, such as during an Electric Power Research Institute (EPRI) conference on June 20, 2023 (Enclosure 2), highlight that when using Code Case N-752 the organization is not required to comply with Section XI, the original construction code and the quality assurance manual. As supported by the EPRI presentation in Enclosure 2 these include reductions in procurement requirements for components, particularly since the material suppliers are not required to be qualified to the QAPM, and reduced quality controls (i.e., non-destructive examinations and pressure testing) since examinations may be visual in lieu of volumetric or surface examinations.

The NRC staff has not endorsed Code Case N-752 in Regulatory Guide (RG) 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, and 10 CFR 50.55a. This is based on concerns raised by NRR technical staff regarding the code case potentially bypassing the Commission requirement to obtain a license amendment for implementation of 10 CFR 50.69, which allows alternative treatment of safety-related LSS components in lieu of 10 CFR Part 50 special treatment requirements, such as specific repair/replacement requirements and Appendix B requirements.

The regulations in 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, state, in part, that ASME BPV Code Class 1, 2, and 3 components must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the Section XI of editions and addenda of the ASME BPV Code and that are incorporated by reference. Therefore, the NRC regulations require that Class 1, 2, and 3 components must meet the requirements of Section XI.

The regulations in Section 50.55a(z), Alternatives to codes and standards requirements, of 10 CFR state, in part, that alternatives to the requirements of 10 CFR 50.55a(b) through (h) of this section or portions thereof may be used, when authorized by the Director, Office of Nuclear Reactor Regulation. An alternative must be submitted and authorized prior to implementation.

The licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, this regulation allows an alternative to the requirements of Section XI above if the alternative provides an acceptable level of quality and safety, or results in a hardship. The differing staff note that, typically, all licensee requests for an alternative to requirements in Section XI are related to specific, individual requirements within Section XI and specific details are submitted regarding what alternative requirements will be followed in order to determine that these alternative requirements provide an acceptable level of quality and safety.

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 alternative stated, The categorization and treatment requirements of Code Case N-752 are consistent with those in 10 CFR 50.69. 10 CFR 50.69 allows removal of all regulatory requirements for repair/replacement, inservice inspection, inservice testing and quality control for LSS structures, systems, and components (SSCs). The differing staff note that Code Case N-752 appears to only remove the repair/replacement requirements. However, plants have implemented risk-informed inservice inspection (RI-ISI) that provides reduced

3 inspection of these components. In addition, licensees have removed the applicability of Appendix B quality assurance requirements from components categorized using ASME Code Case N-752, as Footnote 1 of ASME Code Case N-752 requires the licensee to be exempt from Appendix B requirements in order to fully implement ASME Code Case N-752, as was requested by the ONS proposed alternative. When looking at each program separately, it seems reasonable to reduce the requirements for LSS items, since the other program aspects still provide some assurance of the items ability to perform their functions. As explained above, the use of all of the programs together (Code Case N-752, RI-ISI, removal of Appendix B quality control, etc.), reduces the requirements significantly and approaches the level of reduction allowed in 10 CFR 50.69, except for inservice testing, without the same level of regulatory controls (i.e., license amendment process) and categorization.

This major reduction in requirements in all of these areas, without the specific categorization process similar to 10 CFR 50.69, which includes system categorization (not on component or subcomponent) and other factors such as active/passive functions and not crediting redundant systems with alternative treatment applied as discussed in Concern B of this DPO, does not provide reasonable assurance that the items will be able to perform their functions. Therefore, since the Code Case N-752 categorization process is not the same as or as rigorous as the categorization process in 10 CFR 50.69 (as discussed in Concern B below), there should be consideration of the effects it will have on defense-in-depth. For example, if repair/replacement requirements are reduced, assurance can be provided for the items ability to perform its active and passive functions of the system by monitoring (inservice inspection) and quality assurance of the items materials and construction. However, when all areas are reduced, the Code Case N-752 method is lacking the tools to provide assurance that the subcomponent is able to perform its system safety function. This can be correlated to the five Principles of risk-informed regulations, which consist of:

1. continue to meet current regulations,

2. maintains defense-in-depth,

3. maintains safety margins,

4. cumulative effects are small and do not exceed NRC safety goals.

5. implementation and monitoring strategies which address uncertainties and proposed increase in risk and Of specific note for this case are the principles of defense-in-depth and safety margin. Section 2.1.2, Safety Margin, of RG 1.174 states:

With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met and (2) safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met or proposed revisions provide sufficient margin to account for uncertainty in the analysis and data Use of ASME Code Case N-752 would allow the licensee to use any nationally-approved code or standard as an alternative to the ASME Code. Further the alternative will allow the future reduction of even those standards to any Owners requirements for non-destructive examination and will eliminate the regulatory requirement for pressure testing to verify the integrity of a repair. The differing staff find that, without specific information regarding the alternative codes and standards which may be used and/or what additional changes the owner may implement to those standards, the impact on safety margins and defense-in-depth cannot be determined

4 consistent with the requirement of (z)(1) for the alternative to ensure an acceptable level of quality and safety. The differing staff find that allowing other codes and standards that are further reduced by unknown Owners requirements results in uncertainties in evaluating such a generic and open-ended condition and could have significant changes in safety margin that would lead to questioning the long-term impact on quality and safety under 10 CFR 50.55(a)(z)(1).

Further under Section 2.6, Integrated Decisionmaking, of RG 1.174 states:

In making a regulatory decision, risk insights (including their associated uncertainties) are integrated with considerations of defense in depth and safety margins. The degree to which the risk insights (and their uncertainties) play a role, and therefore the need for detailed staff review, depends on the application.

As noted in NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, the principles of risk informed decision-making can be illustrated as show below in NUREG-1855, Figure 2-1, Elements of the integrated risk-informed decision making process.

Note the highlighted key role of Assess Impact on defense-in-depth and safety margin in the figure above. Further consider the ability to Assess Uncertainties when defining the implementation and monitoring program. NUREG-1855 states that Element 3 in the figure above, Perform a Risk-Informed Analysis, has both a deterministic and probabilistic analysis, as explained in the following.

In this element, an assessment is made, in terms of a risk-informed analysis, to demonstrate that Principles 2, 3, and 4 are met. The risk-informed analysis includes both deterministic and probabilistic components. Appropriate consideration of the uncertainty in both deterministic and probabilistic assessments is required to properly interpret the results. Both the deterministic and probabilistic components implement Principles 2 and 3, which take into account the impact on defense-in-depth and on safety margins. The probabilistic component implements Principle 4, acceptable risk impact. A treatment of the uncertainties in the probabilistic analysis is implicitly required to implement Principles 2, 3, and 4 of risk-informed decision making. Treatment of probabilistic analysis uncertainties is the focus of NUREG-1855.

5 The differing staff believe the decision to move forward with the alternative at ONS was based on evaluation of Principle 4 and a probabilistic risk analysis number without sufficient consideration of the other principles and issues raised by the technical staff through the review of the alternative treatment options of ASME Code Case N-752. Reg Guide 1.174 reinforces the need to use an integrated approach in stating the following.

None of the individual analyses is sufficient in and of itself. In this way, it can be seen that the decision is not driven solely by the numerical results of the PRA.

These results are one input into the decision making and help in building an overall picture of the risk implications of the proposed licensing basis change.

Therefore, the differing staff find the authorization of the alternative at ONS, as written, to implement ASME Code Case N-752 does not meet the principles of risk-informed regulation in the following areas. First, the items no longer are required to meet the repair/replacement requirements in Section XI (which is required by 10 CFR 50.55a(g)(4)) or Appendix B of 10 CFR Part 50. Second, the defense-in-depth and maintaining safety margins have significant uncertainties due to the lack of quality assurance and lack of specific information regarding the alternative codes and standards which may be used and/or what addition changes the owner may implement to those standards. Third, the reduced inservice inspection provided in RI-ISI may not provide adequate monitoring strategies for these items due to the uncertainties in quality assurance and alternative codes and standards to be used and/or modified by each owner. The differing staff note that the previously approved RI-ISI was acceptable based on the component meeting the applicable ASME BPV Code requirements and the Quality Assurance Standards of Appendix B as part of its defense-in-depth assessment. The removal of these defense-in-depth controls under ASME Code Case N-752, while the process under ASME Code Case N-752 relies upon the RI-ISI program to provide adequate performance monitoring, is a further concern that the N-752 process which now removes previously required defense in depth and performance monitoring, relies too heavily on the risk assessment number in performing a risk-informed decision process. Therefore, the differing staff recommend that the use of Code Case N-752 should be looked at in a more integrated manner that supports the principles of risk-informed regulations.

10 CFR 50.55a(z) allows authorization of an alternative to the requirements of Section XI or III of the ASME BPV Code. Each alternative is authorized on the applicability and merits of the application. Approval of the use of one alternative does not constitute approval of that alternative to other applicants or generic approval. Therefore, approving the alternative for ONS on the basis that the past precedent of approving it for ANO is not correct, particularly with the new information gained from industry on the use of Code Case N-752 which includes:

the alternative treatment allows any code or standard (whether applicable or not) and the ability to modify/delete requirements of that standard or use parts of various standards to develop an owners requirement, including specific allowed reduction of NDE and removal of pressure testing requirements regardless of what standard is used.

can designate a component or piece/item of a component (subcomponent) as LSS although the Commission approved methodology would categorize the overall component as HSS and take credit for redundant components/systems that have alternative treatments applied (ability to change all redundant systems to LSS with reduced treatment).

not evaluating the effects of the passive subcomponent on the active function of the whole component.

6 CONCERNS A. Risk-Based vs. Risk-Informed and the Need for Code and Standard Requirements The DPO on the plant-specific use of Code Case N-752 at ONS (ML23262A967) is that it relies too heavily on risk assessment methodologies without adequate defense-in-depth and performance monitoring options, and therefore provides an incomplete assessment of overall safety margin. The differing staff find that this process, used to approve the plant-specific use of Code Case N-752, does not meet the evaluation standard of 10 CFR 50.55a(z)(1), as the finding of an acceptable level of quality and safety cannot be made.

As examples of the NRC staffs previous use of 10 CFR 50.55a(z)(1), the NRC approved the use of ASME Code Case N-660, the predecessor to Code Case N-752, which had specific conditions to define defense-in-depth and did not allow the complete removal of nondestructive evaluation and pressure testing requirements for the repair or replacement of SSCs.

Additionally, risk-informed inservice inspection (RI-ISI) utilizes a similar risk assessment methodology to Code Case N-752, however, its acceptance is based upon maintaining quality assurance requirements and sample inspections to address the defense-in-depth and performance monitoring aspects of a risk-informed decision per RG 1.174, neither of which is included in Code Case N-752. In fact, Code Case N-752 removes the requirement to utilize a post-construction national code or standard in paragraph -1400(f), the requirement for NDE in paragraph -1400(g) and the requirement for pressure testing in paragraph -1400(h) from ongoing RI-ISI programs without additional consideration. This could result in allowing safety-related SSCs to have less requirements than non-safety-related SSCs, in that power plant components are typically required to follow an appropriate nationally approved code and standard with an Authorized Inservice Inspector review.

The differing staff found these concerns directly impacted the quality of the alternative treatment options in the Duke Energy proposal for repair/replacement activities. The differing 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 specific nationally approved codes and standards in their proposed alternative. Duke Energys response did not provide specific alternative codes and standards, but instead reiterated that the basis for ensuring safety was to rely upon the licensees stated commitment to provide reasonable confidence that each LSS item remains capable of performing its safety-related functions under design-basis conditions.

This is the technical basis in the NRC safety evaluation authorizing the alternative. However, again, the differing staff explain that the defense-in-depth allowance of following a nationally-approved code and standard cannot be ensured due to the language of paragraph -1400(f).

The performance monitoring cannot not be ensured due to the allowances of paragraphs -1400(g) and -1400(h) to remove regulatory requirements to follow the nationally approved codes and standards for NDE and pressure testing to ensure structural integrity and quality control of any repair/replacement activity and further use of RI-ISI which eliminates future inspections due to low safety significance risk assessments, which previously assumed meeting national codes and standards. Therefore, the differing staff cannot find that the licensees proposed alternative provides for an adequate level of quality and safety. The differing 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 with clear inclusion under a 10 CFR Part 50, Appendix B program.

7 Based on the uncertainties raised by the lack of definition of what national standards the licensee may opt to use, what additional modifications the licensee may opt to make to the requirements of those national standards, what level of quality assurance activities the licensee may actually implement, etc., associated with the requested alternative, the differing staff cannot make the necessary affirmative finding with sufficient confidence in accordance with 10CFR 50.55a(z)(1) that the alternative will provide for an acceptable level of quality and safety.

Further details of the differing staff review can be found in Enclosures III, IV and V.

The differing staff note that to address the NC on the ONS proposed alternative, DORL management provided a response in Part C of NCP-2023-005 with a final assessment to go forward with approval of the ONS proposed alternative to use Code Case N-752. In Enclosure I, the differing staff note the limitations in the DORL management response to the NC as they relate to the risk informed decision making process.

B.

Passive Function Vs. Active Function Another concern is categorizing a subcomponent of an SSC as LSS for its passive function when the component that it is a part of performs an HSS active function. The safety concern is the impact the relaxed treatment requirements may have on a multitude of subcomponents relating to the passive function and how it may affect the performance of the active function.

A component relies on each part of the component (i.e., subcomponent) to perform its function. Failure of a passive part can affect the active function 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 damage the valve disc (active) and prevent proper opening or closing of the valve.

o Valve body (passive) could fail and divert flow through the valve or prevent the ability to isolate flow by closing the valve.

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

If the valve bonnet is classified as an LSS subcomponent, 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 subcomponent of a component would be extremely difficult to control safely. For example, an MOV actuator has over 100 subcomponents. It would be difficult to determine whether the passive function of each of those subcomponents might have an impact on the active function of the actuator. Code Case N-752 would allow a licensee to classify subcomponents of an MOV actuator to be passive based on their structural integrity. However, many of the internal subcomponents of an MOV actuator can impact the active function of the actuator because of their potential to interfere with the motion of numerous moving subcomponents in the actuator. For example, an improper thermal coefficient of expansion of a replacement subcomponent classified as passive can cause failure of the

8 MOV actuator as a result of interference among the internal subcomponents due to heat-up during valve operation.

In order to determine whether the alternative treatment requirements provide reasonable confidence that the component will perform its function, the specific alternate treatment requirements should be specified and used in their entirety, should be applicable to the specific component, and should be taken into consideration in the categorization process. Code Case N-752 only states that nationally recognized codes and standards (e.g., ASME PCC-2, API-653) may be used. However, Code Case N-752 allows the Owner to modify the requirements of the code or standard as they deem appropriate.

The use of a factor of three to address the additional failure probability of a degraded component or a component that has lower quality is very subjective. If a component has a 10% chance of failure, then increasing that percentage chance to 30% for a component that has lower quality would seem appropriate. However, when a component is assumed to essentially never fail, i.e. 1x10-8, and a degradation or lack of quality is introduced, the direct use of a factor of three will never have an impact on the failure rate and does not accurately account for the lower quality level. Further, it may be difficult to predict, or bound, the overall effect on all modes of failure due to a change in various quality requirements. Using an arbitrary factor of three or even a factor of 10 as a bounding condition for all cases may not provide a reasonable evaluation of the contribution that lower quality or potentially increased degradation probability will have on the failure of a component.

Also, the assumed rate may not be relevant given other information. As an example, the NRC staff found that the quarterly stroke-time testing required by the ASME Code was insufficient to determine the capability of MOVs to perform their safety functions. In NUREG/CR-5140, Brookhaven National Laboratory determined from the results of the MOV testing performed in accordance with NRC Bulletin 85-03 that the increase in failure rate of MOVs could be almost 100 times higher than the assumed failure rate based on data from ASME Code testing.

The differing staff are also taking into account operational experience with the use of Code Case N-752. During a public ASME Code meeting, staff from Arkansas Nuclear One discussed using both categorization processes in 10 CFR 50.69 and Code Case N-752. The licensee highlighted specific instances when their application of Code Case N-752 would lead to significantly different results than their 10 CFR 50.69 methodology. For example, it was stated for the 10 CFR 50.69 evaluation, the emergency feedwater system was high safety significant so the applicable requirements could not be reduced. However, using the Code Case N-752, specific items for these components could be categorized as low safety significance.

This inconsistency is primarily due to the categorization of individual items (such as subcomponents) as LSS rather than categorizing entire systems or structures, contrary to the requirements in 10 CFR 50.69(c)(1)(v) that categorization be performed for entire systems and structures, not for selected components within a system or structure. In addition, Code Case N-752 allows credit for functions of SSCs that have been proposed as LSS or for identical redundant SSCs within the system that are classified as LSS. This can lead to potential miscategorization of an HSS SSC to LSS. Such miscategorization can result in less regulatory treatment being applied to SSCs important to safety. This has the potential to increase the likelihood of failures going undetected for extended periods of time for these HSS SSCs that have been categorized as LSS per Code Case N-752. The methodology in Code Case N-752 also contradicts current methodology of section 6 of NEI 00-04, 10 CFR 50.69 SSC

9 Categorization Guideline, Revision 0 which excludes like components for remaining capability in defense-in-depth categorization (i.e., it does not allow credit for any identical, redundant SSCs within the system that are classified as LSS and have alternative treatment applied). This categorization process of Code Case N-752 also has the potential to overlook the impact of common cause failure interactions, which is inconsistent with section 2.1.1.2, consideration 4, Preserve adequate defense against potential CCFs [Common Cause Failures], of RG 1.174.

Section 2.1.1.3, Evaluating the Impact of the Proposed Licensing Basis Change on Defense in Depth, of RG 1.174 for defense-in-depth consideration 1 (preserve a reasonable balance among the layers of defense), provides additional guidance that states, in part:

...however, to address the unknown and unforeseen failure mechanisms or phenomena, the licensees evaluation of this defense-in-depth consideration should also address insights based on traditional engineering approaches.

Results and insights of the risk assessment might be used to support the conclusion; however, the results and insights of the risk assessment should not be the only basis for justifying that this defense-in-depth consideration is met.

The licensee should consider the impact of the proposed licensing basis change on each of the layers of defense.

As an example, based on the EPRI presentation (Enclosure II), a component may only receive a visual inspection in lieu of volumetric inspection, and therefore flaws could be present within most of the material since only the surface is examined visually. The likelihood of flaws is greater when only a visual examination of the surface is performed in lieu of examining the entire volume of the weld and therefore a greater likelihood of deterioration and failure. Further, a welder knowing a weld is not going to be volumetrically examined, or even a liquid penetrant surface examined, is not going to maintain the same level of quality. Also, new material (different grades of HDPE, carbon fiber repair, etc., that are specifically not approved by the ASME BPV Code) that can be used with no operating experience for nuclear applications could increase the rate of deterioration/failure. In addition, if these alternate treatment repair/replacements are applied to all of the redundant trains, this may compromise the defense-in-depth redundancy of these systems. Therefore, Code Case N-752 should not credit these alternative-treatment repaired systems in the categorization process.

Therefore, the removal of repair/replacement requirements of Section XI based on the risk categorization methodology in Code Case N-752 should consider the effects it will have on defense-in-depth, particularly when these defense-in-depth areas have already been reduced by risk-informed inservice inspection, reduced quality assurance and undetermined code and standard requirements. Removing all of the defense-in-depth criteria does not provide assurance that the components will perform their intended function, and therefore a balance between risk and the defense-in-depth principles would be necessary. In order to provide this balance, the differing staff suggest that the categorization process in Code Case N-752 should not credit any identical, redundant SSCs within the system that are classified as LSS and have alternative treatment applied. This will ensure that components that are categorized as LSS due to redundancy, are not relied upon during the categorization process because some or all of these LSS components can have reduced requirements that are not factored into the categorization process. In addition, the code case should consider including an NEI 00-04 assessment, including Section 10.1, of all active function SSCs that could be affected by the failure of the passive SSC on which a licensee intends to utilize the Code Case N-752 categorization process.

Further basis can be found in the technical concerns section of Enclosure V.

10 C. Removal of the Quality Assurance Requirements The NRC regulations in 10 CFR 50.54, Conditions of licenses, paragraph (a)(4) require that Changes to the quality assurance [QA] program description that do reduce the commitments must be submitted to the NRC and receive NRC approval prior to implementation. In addition, 10 CFR 50.54(a)(3)(ii) indicates that the following is not considered to be a reduction in commitment: 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. Therefore, based on 10 CFR 50.54, if a licensee receives NRC approval of a change to its QA program, another licensee can use that approval per 10 CFR 50.54(a)(3)(ii) to implement a similar change to its QA program, with no further approval from the NRC. Therefore, other nuclear power plant licensees, such as Duke Energy for ONS, are relying on 10 CFR 50.54(a)(3)(ii) to implement the same QAPM changes as Entergy made at ANO to remove safety-related Class 2 and Class 3 SSCs categorized as LSS from the scope of their 10 CFR Part 50, Appendix B, program when implementing Code Case N-752 at their nuclear power plants without a requirement to submit those changes for NRC review.

Entergy submitted a request to reduce its commitments in the Quality Assurance Program Manual (QAPM) at ANO during implementation of Code Case N-752 under 10 CFR 50.54 on October 26, 2020, as supplemented by letters dated April 5, 2021, and April 30, 2021. On May 19, 2021 (ML21132A279), the NRC staff issued an SE approving the proposed change to the QAPM at ANO under 10 CFR 50.54 for safety-related Class 2 and 3 components categorized as LSS when implementing Code Case N-752.

According to its statements in those submittals and to NRC Regional inspection staff, Entergy considers the NRC SE to have granted an exemption from the QA requirements in 10 CFR Part 50, Appendix B, for safety-related Class 2 and Class 3 items categorized as LSS when implementing Code Case N-752 at ANO. The scope of Code Case N-752 includes a multitude of safety-related Class 2 and Class 3 items categorized as LSS that could receive reduced QA treatment under Code Case N-752 with the QAPM changes over the remaining life of ANO. Such reduced QA treatment over an extended time could lead to common cause failures of two or more of these items to perform their active or passive safety-related functions that might result in a significant impact on the safe operation of ANO. As indicated by changes to its QAPM and SAR, Entergy assumes that the licensing bases and enforcement actions related to safety-related Class 2 and Class 3 items categorized as LSS when implementing Code Case N-752 at ANO are outside the scope of 10 CFR Part 50, Appendix B.

Therefore, the differing staff for that issue requested formal initiation of a DPO on the NRC SE granting the ANO QAPM changes.

The differing staff note that Duke Energy has utilized 10 CFR 50.54(a)(3)(ii), as a basis, to claim this same perceived exemption as Entergy believes they have obtained. The differing staff note that neither Entergy nor Duke Energy has requested an 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 differing staff believe such an exemption is necessary prior to implementing the proposed alternative or Duke Energy should modify their alternative. Based on the above, the SE authorizing the use of ASME Code Case N-752 at ONS under 10 CFR 50.55a(z)(1) violates the NRC rules for granting exemptions from the NRC regulations in light of information

11 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.

Based on the significance of the NRC decision, the Duke Energy request to apply Code Case N-752 as an alternative under 10 CFR 50.55a(z)(1) should be reviewed by OGC. In addition, the ONS alternative that was authorized by the NRC SE should include all items classified as LSS under Code Case N-752 in their Appendix B program but may reduce the level of rigor for these items with reasonable assurance that each LSS item remains capable of performing its safety-related functions under design-basis conditions.

On April 10, 2024 (ML24101A388), Entergy submitted a response to an NRC request for confirmation of information (RCI) related to its alternative request to implement Code Case N-752 at several other nuclear power plants. In the RCI response, Entergy uses phrases such as are being treated and currently implemented in discussing its application of procedures and processes that were developed in accordance the provisions of 10 CFR Part 50, Appendix B, for the repair and replacement of safety-related Class 2 and Class 3 SSCs categorized as LSS when implementing Code Case N-752. The licensee did not specifically state that 10 CFR Part 50, Appendix B applied to activities performed using these processes and procedures for components under Code Case N-752. Therefore, Entergy is indicating its view that the application of 10 CFR Part 50, Appendix B, is voluntary for repair and replacement of safety-related Class 2 and Class 3 SSCs categorized as LSS when implementing Code Case N-752. Entergy has not retracted its statements in its earlier submittals that 10 CFR Part 50, Appendix B, is not applicable to safety-related Class 2 and Class 3 SSCs categorized as LSS when implementing Code Case N-752 at the ANO nuclear power plant. NRC safety evaluations cannot impose Appendix B to 10 CFR Part 50 at any nuclear power plant where a licensee references the ANO submittals when implementing Code Case N-752, because an NRC safety evaluation cannot impose a condition that is contrary to statements in a licensees submittal.

Further details of the differing staff review can be found in Enclosure IV.

D. Legal authority 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. 10 CFR 50.69 is the regulation that allows removal of all regulatory requirements for repair/replacement and quality control for safety-related LSS SSCs. This also includes the policy issue of whether approving alternative requests for using ASME Code Case N-752 deviate from the Commission direction that a license amendment is needed to apply risk-informed categorization per 10 CFR 50.69, and whether the Code Case allows a licensee to be exempt from other portions of the NRC regulations, such as in 10 CFR Part 50, Appendix B.

12 Therefore, the safety evaluation for ONS 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, and for future plants.

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

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

However, since the code case has not been generically endorsed, it is not in scope with the DPO process and therefore is not included as part of this DPO.

SUGGESTED RESOLUTION

1. Licensees should include all items classified as LSS under Code Case N-752 in their Appendix B program but may reduce the level of rigor for these items with reasonable assurance that each LSS item remains capable of performing its safety-related functions under design-basis conditions.

2. The licensee will establish minimum requirements for specific codes and standards for repair/replacement activities associated with the implementation of the plant-specific proposed alternative using Code Case N-752.

3. Request OGC review 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 Code Case N-752 at ONS:

a) Confirm that the Staff cannot authorize the use of 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, or a separate license amendment in accordance with 10 CFR 50.90.

13 b) 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.

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

4. The categorization process in Code Case N-752 should exclude like components for remaining capability in defense-in-depth categorization (i.e., it does not allow credit for any identical, redundant SSCs within the system that are classified as LSS and have alternative treatment applied). In addition, the code case should consider including an NEI 00-04 assessment, including Section 10.1, of all active function SSCs that could be affected by the failure of the passive SSC on which a licensee intends to utilize the Code Case N-752 categorization process.

SUPPORTING INFORMATION See Enclosures.

CONCLUSION The authorization to use Code Case N-752 at ONS is based on risk-based principles. The code case lacks sufficient defense-in-depth principles and compromises safety margins due to the lack of quality assurance and a definitive code or standard to be applied to the item. In addition, since there are no minimum requirements (i.e., specific alternative codes and standards) in the code case, or the NRC approval, there is insufficient basis for determining whether the requirements provide an acceptable level of quality and safety as required by 10 CFR 50.55(a)(z). The NRC legal authority to approve such a relaxation of the ASME Code contradicts and bypasses a similar process already specified in the regulations (10 CFR 50.69).

The differing staff believe the NRC authorized ONS alternative to utilize ASME Code Case N-752, with no deviations, is a risk-based approach, relying heavily on the probabilistic risk assessment. The alternative treatment controls of relying on promises of the licensee to maintain safety with no regulatory or quality controls provides no additional value to the risk informed decision process.

14 ENCLOSURE I LIMITATIONS IDENTIFIED IN DISPOSITION OF ASSOCIATED NON-CONCURRENCE (NC)

The differing staff note that in response to the NC on the ONS proposed alternative, DORL management provided a response in Part C with the final assessment to go forward with approval of the ONS proposed alternative to use Code Case N-752. The differing staff provide the following comments regarding the DORL management responses to the three issues in the NC.

Issue 1 - Quality Assurance Program and Appendix B a) Concerning the following management response that states:

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.

This statement is irrelevant since per the regulations in 10 CFR 50.54(a)(3)(ii), Duke Energy can use the previous submittal for ANO to be exempt from the NRC regulations.

As indicated by changes to the ANO QAPM and SAR, Entergy assumes that the licensing bases and enforcement actions related to safety-related Class 2 and Class 3 items categorized as LSS when implementing Code Case N-752 at ANO are outside the scope of 10 CFR Part 50, Appendix B. Therefore, other nuclear power plant licensees, such as Duke Energy for ONS will rely on 10 CFR 50.54(a)(3)(ii) to implement the same QAPM changes as Entergy made at ANO to achieve an exemption from 10 CFR Part 50, Appendix B, without a requirement to submit those changes for NRC review.

b) Concerning the following management response that states:

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.

The Entergy submittal dated 10-26-2020 (ML20300A324) states that LSS components are outside the scope of Appendix B. For example, In Note 5 of Table 1, Entergy states the following:

Treatment (e.g., design control, configuration control, procurement, installation) of Class 2 and 3 LSS items will not be required to comply with the quality assurance provisions of 10 CFR 50, Appendix B. However, the procedures governing these treatment activities will be classified as safety-related and therefore, under the jurisdiction of 10 CFR 50, Appendix B.

15 These statements by Entergy reflect its incorrect assumption that 10 CFR Part 50, Appendix B, will not apply to safety-related Class 2 and Class 3 SSCs categorized as LSS when implementing Code Case N-752 at ANO. Entergy has revised its QAPD to remove LSS components. Therefore, the NRC SE on the ANO use of Code Case N-752 is invalid and is currently being used by ONS and proposed by other licensees.

c) Concerning the following management response that states:

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.

The use of reasonable confidence in 10 CFR 50.69 applies to LSS treatment that is not required to meet Appendix B by a license amendment. Without an exemption or license amendment to apply 10 CFR 50.69, licensees are required to meet Appendix B.

Therefore, the Duke Energy needs to define reasonable confidence in a manner that it meets Appendix B when implementing Code Case N-752.

Issue 2 - Codes and Standards a) Concerning the following management response that states:

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.

The DORL management response indicates that the differing staff did not identify a safety concern. The differing staff identified several safety concerns, including operating experience.

i. NRC Regional Staff provided input from ANO to identify current ongoing concerns regarding use of ASME Code Case N-752.

ii. Previous MOV operational experience in Bulletin 85-03 and NUREG/CR-5140 that showed a high increase of valve failure rate from 0.003% to 8.4% due to inadequate activities performed on the valves such as testing. This could be corelated directly to the Code Case N-752 alternative treatment of similar passive components. This rate of failure increase was not evaluated in the supplemental risk assessment of this concern.

iii. Examples of operating experience of passive components affecting the capabilities of the active components, such as bonnets (passive component) affecting the stem (active component) of MOVs.

iv. The safety concern that when these alternative treatment repairs are placed back in service, the failure probabilities for these items have significant uncertainty due to the treatment allowances of ASME Code Case N-752. Licensees will then utilize the previous data from fully regulated and quality assured repair/replacement activities or original quality-controlled equipment to determine failure probabilities for this

16 equipment. This increase in uncertainty will continue to rise as additional repairs are performed through the remaining years of operation.

In accessing these safety concerns, the NC staff noted that the DRA staff performed a risk assessment to evaluate the concern of reduced passive component quality controls future impact on active components in order to address some of the concerns in Item 1 above. They identified the risk as non-negligible and within Region II of Figure 4 in Regulatory Guide 1.174. Therefore, additional consideration should be applied in the risk informed decision process. This is considered a safety concern.

The differing staff further note that the DRA staff evaluation was based on a maximum factor of 10 increase in valve failure rate to justify that the overall risk assessment for passive components effecting active components as considered low. Unfortunately, as the differing staff have explained, this factor does not address all of the concerns in Item 1 above. Of specific note, NUREG/CR-5140 has shown that the valve failure rates could increase by 100 times the historical valve failure rates, thereby indicating that the risk could be much higher.

The differing staff further tried to explain the safety and quality concerns by providing examples of previous fully qualified and inspected components failing and requiring repair. The differing staff provided this information to explain how allowing engineering judgement in lieu of an applicable national code or standard to maintain an acceptable level or quality and safety creates significant uncertainties in how the failure rate might increase. These uncertainties, caused by lack of non-destructive examination and no pressure testing as allowed by Code Case N-752, eliminate significant defense-in-depth and performance monitoring steps of an appropriate national code or standard. Each of these factors cause concern in the uncertainty in evaluating the alternative treatments quality and safety in future applications.

DORL management also found it unnecessary to be so prescriptive in identifying a specific standard to arrive at a reasonable assurance finding, however, Code Case N-752 itself provides readily available and appropriate non-nuclear codes and standards used for power plant systems such as B31.1, ASME PCC-2 and API-653 as optional examples. Requesting the identification of legitimate codes and standards of sufficient quality should be straightforward and not burdensome. For the licensee to have "reasonable confidence" of the categorized LSS components/subcomponents, the specific codes and standards requirements to be used must be known. Each of these standards to be used, when required in whole, would provide a reasonable basis for assuming a minimum level of quality and control was being utilized in a repair/replacement activity. It would reduce the level of the differing staffs concern of uncertainty to a reasonably acceptable margin versus the current allowance of any engineering judgement option for the repair/replacement process, NDE and pressure testing requirements.

Additionally for the codes and standards issue, the DORL management response highlighted SRM-19-0036. SRM-19-0036 consists of a paragraph on the NuScale Design Certification. The differing staff believe the DORL management response is focusing on the sentence, In any licensing review or other regulatory decision, the staff should apply risk-informed principles when strict, prescriptive application of deterministic criteria

17 such as the single failure criterion is unnecessary to provide for reasonable assurance of adequate protection of public health and safety.

The differing staff believes our position is in accordance with the statement. The use of Code Case N-752 to reduce significant regulatory requirements is reasonable based on the evaluation of the risk significance of the SSC. However, as part of the risk-informed decision making process, it must include a process to review the impact of the proposed alternative beyond just the risk-based calculation. Factors such as defense-in-depth and performance monitoring to ensure sufficient safety margin must be addressed.

RG 1.174 states, None of the individual analyses is sufficient in and of itself. In this way, it can be seen that the decision is not driven solely by the numerical results of the PRA.

SRM-19-0036 did not negate the staff process to assess each of the factors in a risk-informed decision. The SRM simply stated that the NRC staff should implement risk-informed principles in accessing the need for deterministic requirements in licensing reviews or other regulatory decisions.

b) Concerning the following management response that states:

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.

DORL management also noted the licensees controls as being adequate. However, there is no regulatory requirement to those controls. Controls which are essentially a promise to ensure each component meets its intended safety function and had not yet been developed by the time of authorization of the alternative, beyond plant walkdowns.

This appears to be an approval based upon a licensee commitment. LIC-105, Managing Regulatory Commitments Made by Licensees to the NRC, states, a licensee is not legally bound to fulfill, and subsequently control, an action appropriately classified as a regulatory commitment." As part of the commitment process, and as specified in Division of Operating Reactor Licensing (DORLs) Regulatory Commitment Training Modules, if the NRC staff needs to rely on a regulatory commitment in an SE, the staff must escalate the commitment to an obligation (such as the license) or incorporate it into a mandated licensing basis document. Therefore, the NRC staff cannot make a licensing decision if that decision is based on the commitment. The differing staff are concerned that reliance on these controls rather than using an appropriate code and standard under an Appendix B controlled process will lead to the only way to assess the potential failure of these controls will be when a component fails to meet its safety function. It was the NC staffs position that a determination of safety cannot be made based only an assumption that the licensee will come up with adequate controls with no regulatory basis to enforce them.

c) Concerning the following management response that states:

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

18 as alternatives, although neither the NRC nor licensee has identified a realistic scenario where this would occur at ONS.

Typically, in the proposed alternative process, the NRC does not evaluate an alternative based on undefined controls that will be defined at some future date. The NRC evaluates a technical basis for what is to be done in order to effectively evaluate the level of quality and safety. It should be noted that the burden is not on the staff to prove that the process will fail by coming up with example scenarios. The burden is on the licensee to prove that the alternative will consistently and reliably provide for an acceptable level of quality and safety. However, the differing staff did provide example scenarios of inappropriately applying standards in Enclosure II of the NCP such as the following:

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. Its also noted that there have been historical examples (e.g., Davis Besse) of licensees implementing bad decisions based on financial cost savings.

Issue 3, Legal Authority a) Concerning the following management response that states:

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.

Regulations are not written to either explicitly or implicitly prohibit the NRC from doing all matters.

b) Concerning the following management response that states:

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

The NCP is not insisting it be done by LAR, but stating fact that the alternative being reviewed explicitly stated the use of the code case that is being requested to be approved are consistent with those in 10 CFR 50.69.

c) Concerning the following management response that states:

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.

This is a policy issue that needs to be established and not by merely stating that the regulations do not explicitly prohibit it. In addition, the alternative requested does not determine the impact on active components, which could have safety implications and legal concerns as discussed in the NCP enclosures. Neither of which is answered in this response.

d) Concerning the following management response that states:

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.

Contrary to this statement, the Entergy submittal dated 10-26-2020 states that LSS items are outside the scope of Appendix B when implementing Code Case N-752.

Therefore, OGC review is needed for the ONS SE to determine path forward in rectifying this contradictory allowance in the Entergy submittal that ONS can implement by 10 CFR 50.54(a)(3)(ii).

e) Concerning the following management response that states:

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.

Contrary to this statement, Duke Energy proposes to remove LSS items from its QA plan which establishes the Appendix B scope at ONS. Based on the ONS submittals and past precedence (ANO), the licensee effectively will remove these LSS items from its Appendix B program contrary to the management response.

20 f) Concerning the following management response that states:

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.

Since the use/non-use of past precedence does not require a backfit analysis, the SE conclusion to approve it based on past approval is not correct, given the new industry information and understanding of the code case that was provided in the NCP. The approval of an alternative request to use a code case with new information that allows one process to categorize a component as HSS, while another process allows the same component to be classified as LSS, should warrant the staff to question the process.

That is what the staff did, and based on the NCP enclosures determined that there are safety implications that should be addressed. This includes that the code case allows crediting for any identical, redundant SSCs within the system that are classified as LSS and have alternative treatments applied, and that an evaluation of the passive function affecting the active function of SSCs are not performed. Past experience includes passive valve bonnets affecting the active function of the valve stems. Therefore, the NRC should be cautious in approving such a request, or future requests, to change repair/replacement activities that have clear safety implications as discussed in the NCP enclosures.

21 ENCLOSURE II.

EXCERPTS FROM EPRI PRESENTATION

22

23

24 ENCLOSURE III.

ALTERNATIVE CODES AND STANDARDS CONCERN PROPOSED SAFETY EVALUATION DENIAL This enclosure provides the differing 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 differing staff cannot make a determination that the licensees proposed alternative provides for an adequate level of quality and safety. Therefore, the differing staff find 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 differing 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 differing 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 differing 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 (ANO) 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 differing staff review.

The differing staff also considered the operating experience of these programs in the review for the plant-specific application at ONS. The differing staff identified a concern related to the ASME Code Case N-752 application versus the 10 CFR 50.69 application, which includes 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. Differing staff then considered the impact of defense-in-depth and safety margin, the second and third principles in RG 1.174, respectively. 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 differing staff performed a focused review on the potential impact of the licensees proposed alternative requirements relating to codes and standards for

25 repair/replacement activities on safety margin. The differing 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 differing 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 and thereby maintaining adequate safety margin in accordance with the risk informed decision making process.

Reasonable Confidence Issue The differing 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 differing 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 differing staff find 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 differing 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 differing 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 differing 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 differing staff again note that the NRC has not generically approved Code Case N-752, therefore the specific language of the licensees submittal was the focus of the differing staff review.

In reviewing the licensees specific alternative treatment wording of Section 5.2.E, the differing 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 an adequate level of quality and safety. The differing staff recognize that the general basis for the proposed alternatives approach was to replace the requirements of Section XI of the ASME Code with requirements from the original Construction Code, Owners

26 Requirements, and nationally recognized codes, standards, or specifications applicable to the LSS categorized item as permitted by the licensing basis. However, the differing staff found the specific language of the following Sections (discussed below) allow the use of Owners options in lieu of specific codes and standards, which causes significant uncertainty in evaluating the safety margin and a concern to the quality of the licensees proposed alternative treatment.

Section 5.2.E.6 of the licensees submittal states the following, The repair methods of nationally recognized post-construction codes and standards (e.g.,

PCC-2, API-653) applicable to the item [may] [emphasis added] be used.

The differing staff note the language would allow the Owner to choose whether or not to apply a nationally recognized post-construction code or standard raising concerns about the quality of the alternative treatment. The repair methods allowed by this language are not clearly established. Further, this lack of clarity would apply for any repair made on an item for the duration of this alternative. The differing staff note that operating experience has identified numerous U.S. nuclear plant Owner alternative repair methods that the NRC has found unacceptable to provide reasonable assurance of plant safety. The differing staff find the allowance of the Owners options for repair methods in ASME Code Case N-752 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 the may wording, stating, The word shall is not used because shall would establish that these nationally recognized post-construction codes and standards must be used. The use of the word may retains the flexibility to use Section XI code, if desired or if there is not a nationally recognized post-construction codes or standard applicable to the item.

The differing staff consider Section XI of the ASME Code to be a nationally recognized post construction code, therefore its use would still be allowed through the changing of the word may to shall. If there is not a nationally recognized post-construction code or standard applicable to the item, then the default is to utilize the original Construction Code, as the original Construction Code is what the licensees license allows for operation with this repaired/replaced safety item. Additionally, post-construction codes have similar citations, such as Section 101-3.11, Examination, on page 3 of ASME PCC-2 which states, When qualifications of examiners, methods of examination, extent of examination, and acceptance criteria are not specified, they should follow the requirements of an appropriate construction code or post-construction code.

The versatility allowed by alternative post-construction codes to the current regulatory requirement of Section XI provides significant flexibility, but they still maintain a requirement for an established specific code or standard. The differing staffs concern remains that if a nationally approved post-construction code is not defined then Owners options to use any alternative, including engineering judgement, for repair methods is allowed by the current wording of Section 5.2.E.6 of the licensees submittal. The differing staff find this 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.

Section 5.2.E.7 of the licensees submittal states the following, in part,

27 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 differing 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 differing staff find 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 differing 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 differing 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 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 differing staff find 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

28 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 differing staff find 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 differing staff also observe 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 differing staff consider 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 differing staff find 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 differing staff note 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 differing staff find 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 differing staff find 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 differing 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.

Consideration of Licensee Programs The differing 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

29 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 differing staff note that the process under 10 CFR 50.55a(z) to authorize the licensees proposed alternative to implement Code Case 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 differing 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 differing staff cannot find that the licensees proposed alternative provides for an adequate level of quality and safety.

The differing 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 differing staff used risk insights to evaluate the overall impact of the use of ASME Code Case N-752. The differing staff note 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 differing staff believe the

30 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 differing 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 differing 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 differing staff note 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 differing staff recognize 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 differing staff find 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 differing staff note 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 differing staff, utilizing risk insights to consider the impact of not requiring nationally approved codes or standards, have 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 differing staff find the licensees proposed alternative does not allow the staff to determine that it will provide for an adequate level of quality and safety. Therefore, the differing staff find the licensees proposed alternative does not meet the requirements for authorization under 10 CFR 50.55a(z)(1) as requested.

The differing staff find 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.

31

32 ENCLOSURE IV.

10 CFR PART 50, APPENDIX B, APPLICABILITY The Division of Operating Reactor Licensing (DORL) in the NRC Office of Nuclear Reactor Regulation (NRR) prepared a safety evaluation (SE) authorizing Duke Energy Carolinas, LLC (Duke Energy) to implement American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code) 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 does not ensure that Duke Energy will be legally required to comply with the Quality Assurance (QA) requirements in 10 CFR Part 50, Appendix B, when performing repair and replacement activities for safety-related Class 2 and 3 items categorized as low safety significant (LSS) 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 Class 2 and 3 items when implementing ASME Code Case N-752. 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.

During preparation of the SE by DORL, the NRR Mechanical Engineering and Inservice Testing Branch (EMIB) provided a 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 use of 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. DORL did not provide the proposed RAI to Duke Energy to clarify that compliance with 10 CFR Part 50, Appendix B, is required for safety-related LSS Class 2 and 3 items when implementing ASME Code Case N-752.

Specific examples where ASME Code Case N-752 and Entergy and Duke Energy documents improperly indicate that the QA requirements of 10 CFR Part 50, Appendix B, do not apply at ANO and ONS for safety-related LSS components when implementing ASME Code Case N-752 are as follows:

(1) ASME Code Case N-752, paragraph -1420, LSS Items, states that LSS items do not have to meet the requirements of IWA-1400(o). However, Footnote 1 of the Code Case states: If compliance with 10 CFR 50 Appendix B or NQA-1 is required at the Owners facility, IWA-1400(o) is not exempt.

The requirements of IWA-1400(o) address the documentation requirements to be applied for the QA program and specifically invoke the requirements of 10 CFR Part 50, Appendix B, and ASME Standard NQA-1, accepted in 10 CFR 50.55a provided it is implemented consistent with Appendix B. Hence, the requirements of IWA-1400(o) are largely specific to the United States (U.S.). When the Code Case was developed, in order to make it internationally applicable, ASME included paragraph -1420 in the Code

33 to make it clear that users outside of the U.S. did not have to meet the U.S.-specific requirements of IWA-1400(o). However, NRC staff participating in the ASME BPV Code process ensured that Footnote 1 was inserted into the Code Case to clarify that requirements imposed on U.S. plants still had to be met.

(2) ASME Code Case N-752, paragraph -1420, states that the Owner shall define requirements (e.g., design control, procurement, installation, configuration control, and corrective action) to confirm with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design-basis conditions.

Although Code Case N-752 does not define reasonable confidence, based on the development of Code Case N-752, the staff understands that the term reasonable confidence is intended to mean the same as it does in 10 CFR 50.69. In 10 CFR 50.69, in order to apply the standard of reasonable confidence, a licensee is expressly exempted from the requirements of 10 CFR Part 50, Appendix B, when the license amendment approving 10 CFR 50.69 is issued. However, no exemption from 10 CFR Part 50, Appendix B, has been submitted or granted to support the implementation of the reasonable confidence standard under Code Case N-752 for either ANO or ONS.

(3) The Entergy submittal dated October 26, 2020 (ML20300A324) for the implementation of Code Case N-752 at ANO in the cover letter states:

Code Case N-752 exempts Class 2 and 3 LSS SSCs from the repair/replacement requirements of ASME Section XI; it also allows exemption of the associated QA requirements [i.e., IWA-1400(n)] provided 10 CFR 50 Appendix B or NQA-1 is not required at the nuclear site.

Table 1 in Note 5 in the Entergy submittal dated October 26, 2020 (ML20300A324) for the implementation of Code Case N-752 at ANO states:

Treatment (e.g., design control, configuration control, procurement, installation) of Class 2 and 3 LSS items will not be required to comply with the quality assurance provisions of 10 CFR 50, Appendix B. However, the procedures governing these treatment activities will be classified as safety-related and therefore, under the jurisdiction of 10 CFR 50, Appendix B.

(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).

34 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.

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.

35 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 not indicated 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. In lieu of 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 the IQVB input did not resolve the inconsistency in the licensee submittals and statements. Further, EMIB was not included in the concurrence chain of the DORL SE although EMIB provided significant participation in the NRC staff review of the Duke Energy request to use Code Case N-752 at ONS.

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.

36 On December 13, 2023, the NRC staff issued an SE authorizing Duke Energy to implement Code Case N-752 under 10 CFR 50.55a(z)(1) at ONS, based on the SE dated May 19, 2021, granting the ANO request. The NRR technical branch responsible for this review (Piping and Head Penetrations Branch - NPHP) and the assigned staff reviewer in NRR EMIB determined that the Duke Energy request was unacceptable based on the lack of requirements for the treatment of safety-related Class 2 and 3 items when implementing Code Case N-752, and the Duke Energy reliance on the ANO submittals removing safety-related Class 2 and 3 items from the scope of Appendix B. Without resolving these safety concerns, NRR management removed the review of the Duke Energy request from the applicable Technical staff and assigned Projects staff to prepare an SE authorizing the request. The NPHP Branch Chief and the assigned technical staff reviewers filed an NCP on the ONS SE prepared by Projects staff that did not resolve the safety concerns.

The ONS SE dated December 13, 2023, in the Quality Assurance section did not resolve the misstatements in the Entergy submittals for ANO. Prior to issuance of the SE, Duke Energy should have been required to retract the statements in the ANO submittals that 10 CFR Part 50, Appendix B, does not apply to safety-related Class 2 and 3 LSS items when implementing Code Case N-752. Further, the SE should have included the following discussion to identify and address the misstatements by Entergy:

On May 27, 2020 (ML20148M343), with a supplement dated November 17, 2020 (ML20322A141), Entergy Operations, Inc. (Entergy) submitted a request to implement ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1, at Arkansas Nuclear One, Units 1 and 2 (ANO) as an alternative to specific requirements in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),

Section XI, under Title 10 of the Code of Federal Regulations (10 CFR) in Section 50.55a, Codes and Standards, paragraph (z)(1), as providing an acceptable level of quality and safety. On October 26, 2020 (ML20300A324), with supplements dated April 5, 2021 (ML21095A244), and April 30, 2021 (ML21120A326), Entergy submitted a request to reduce its commitments in the Quality Assurance Program Manual (QAPM) at ANO during implementation of Code Case N-752 under 10 CFR 50.54, Conditions of licenses. On May 19, 2021 (ML21118B039), the NRC staff issued a safety evaluation (SE) authorizing the use of Code Case N-752 at ANO as an alternative to the specified ASME BPV Code,Section XI, requirements under 10 CFR 50.55a(z)(1). On the same date of May 19, 2021 (ML21132A279), the NRC staff issued an SE approving the proposed change to the QAPM at ANO under 10 CFR 50.54 with specific QA requirements under 10 CFR Part 50, Appendix B (although less rigorous than previously applied at ANO under Appendix B) for safety-related Class 2 and Class 3 components categorized as low safety significant (LSS) when implementing Code Case N-752 at ANO. The NRC approval of the changes to the QAPM at ANO are based on the specific QA requirements for safety-related LSS Class 2 and Class 3 components when implementing Code Case N-752 documented in the Entergy submittals dated October 26, 2020, April 5, 2021, and April 30, 2021. Contrary to the incorrect statements in the Entergy submittal dated October 26, 2020, and other Entergy documents that 10 CFR Part 50, Appendix B, does not apply to Code Case N-752 repair/replacement activities, Entergy is required under 10 CFR Part 50, Appendix B, to meet the QA requirements specified in the Entergy submittals dated October 26, 2020, April 5, 2021, and April 30, 2021, for safety-related LSS Class 2 and Class 3 components when implementing Code Case N-752 at ANO. Further, regulatory licensing basis and enforcement actions related to 10 CFR Part 50, Appendix B, continue to apply to safety-related LSS Class 2 and Class 3 components when implementing Code Case N-752 at ANO.

37 Additional nuclear power plant licensees are rapidly proposing to implement Code Case N-752 based on the ANO SE. For example, NextEra submitted a request on March 15, 2023, under 10 CFR 50.55a(z)(1) to implement Code Case N-752 at St. Lucie Units 1 and 2, Turkey Point Units 3 and 4, Seabrook Station, and Point Beach Units 1 and 2, relying on the ANO SE as the precedent for its request.

On April 10, 2024 (ML24101A388), Entergy submitted a response to an NRC request for confirmation of information (RCI) related to its alternative request to implement Code Case N-752 at several other nuclear power plants. In the RCI response, Entergy uses phrases such as are being treated and currently implemented in discussing its application of procedures and processes that were developed in accordance the provisions of 10 CFR Part 50, Appendix B, for the repair and replacement of safety-related Class 2 and Class 3 SSCs categorized as LSS when implementing Code Case N-752. The licensee did not state specifically that 10 CFR Part 50, Appendix B applies to activities performed using these processes and procedures for components under Code Case N-752. Therefore, Entergy is indicating its view that the application of 10 CFR Part 50, Appendix B, is voluntary for repair and replacement of safety-related Class 2 and Class 3 SSCs categorized as LSS when implementing Code Case N-752. Entergy has not retracted its statements in its earlier submittals (such as the ANO submittal dated October 26, 2020) that 10 CFR Part 50, Appendix B, is not applicable to safety-related Class 2 and Class 3 SSCs categorized as LSS when implementing Code Case N-752 at the ANO nuclear power plant. NRC safety evaluations cannot impose Appendix B to 10 CFR Part 50 at any nuclear power plant where a licensee references the ANO submittals when implementing Code Case N-752, because an NRC safety evaluation cannot impose a condition that is contrary to statements in a licensees submittal.

Based on the above facts, the DORL SE authorizing the use of ASME Code Case N-752 at ONS under 10 CFR 50.55a(z)(1) could be interpreted as violating 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 applied to safety-related-LSS components when implementing Code Case N-752.

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. As a result of the Entergy statements that Appendix B does not apply to safety-related LSS Class 2 and Class 3 components when implementing Code Case N-752 at ANO, a DPO has been submitted requesting correction of the NRC SE dated May 19, 2021, granting the revision of the QAPM at ANO.

38 ENCLOSURE V.

BACKGROUND FOR OGC QUESTIONS 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors 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 allow 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 with high level treatment requirements for low risk classified SSCs. A licensee voluntarily choosing to implement this section shall submit an application for a 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 (safety significant safety-related) SSCs, RISC-2 (safety significant non-safety-related) SSCs, RISC-3 (low safety significant safety-related) SSCs, and RISC-4 (low safety significant non-safety-related) 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:

39 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

40 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.

10 CFR 50.69 FRN (69 FR 68019, 68039) states the Commissions intent that 50.69 RISC classification would be performed at the component level as follows:

This required scope ensures that all safety functions associated with a system or structure are properly identified and evaluated when determining the safety significance of individual components within a system or structure and that the entire set of components that comprise a system or structure are considered and addressed.

Additional information is also provided in the ASME Code, 10 CFR 50.55a and applicable NRC RGs. 10 CFR 50.55a provides rules for SSCs classified as Quality Groups A, B, and C corresponding to ASME Code Class 1, 2 and 3 and states:

Guidance for quality group classifications of components that are to be included in the safety analysis reports pursuant to § 50.34(a) and § 50.34(b) may be found in Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radiological-Waste-Containing Components of Nuclear Power Plants, and in Section 3.2.2 of NUREG-0800, Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants.

Paragraph IWA-1320 of Section XI of the ASME BPV Code, Rules for Inspection and Testing of Components of Light-Water-Cooled Plants, states, Application of the rules of this Division shall be governed by the group classification criteria of the regulatory authority having jurisdiction at the plant site as follows. RG 1.26, Quality Group Classifications and Standards For Water-,

Steam-, And Radioactive-Waste-Containing Components Of Nuclear Power Plants, provides one method of classifying SSCs and also includes that 10 CFR 50.69 may also be used for classification. In addition, Appendix A to RG 1.26 provides additional information concerning use of 10 CFR 50.69 as it relates to the classification, including the following:

The applicant or licensee should have implementing procedures for properly categorizing each component using 10 CFR 50.69. The plant procedures should be consistent with the NRC-approved categorization process as described in the applicable final safety analysis report and sufficiently detailed to provide assurance of proper categorization of components. The description of the categorization of SSCs into RISC-1, RISC-2, RISC-3, and RISC-4 categories should include the process to categorize the safety significance of components based on the active (mechanical and electrical) functions of a component, the passive functions of a component (pressure boundary), and, for those components that are modeled in the probabilistic risk assessment, the importance of the component to the risk estimates.

As stated in Section XI of the ASME BPV Code, classification is governed by the regulations of each regulatory authority.

ASME Code Case N-752

41 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.

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 proposed changes to ASME Code Case N-752-1, which were incorporated into Code Case N-752-2, based on the pilot ANO relief request, lessons learned from the pilot, and related industry efforts.

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.

42 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.

However, 10 CFR 50.54(a)(4)(ii) requires changes to the quality assurance program description that do reduce commitments must be submitted to the NRC for approval prior to implementation. The submittal must include all pages affected by that change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to satisfy the criteria of Appendix B of this part. This appears to require that any deviations from the quality assurance program must still continue to meet Appendix B.

The staff previously approved (per safety evaluation ML21132A279 for ANO) a change to the treatment of safety-related Class 2 and 3 SSCs identified as LSS in accordance with the ASME Code Case N-752 to not be required to meet the requirements of the QAPM in accordance with 10 CFR 50.54(a)(4). Instead, the licensee would develop program elements describing treatment of these LSS SSCs to ensure continued capability and reliability of the design-basis function.

Inconsistencies Between 10 CFR 50.69 and ASME Code Case N-752 As noted earlier, the differing staff has concerns with the generic applicability of ASME Code Case N-752. For example, the Code Case appears inconsistent with the concepts and intent of 10 CFR 50.69 in several aspects. Some of these aspects are highlighted below:

Definitions Item in Code Case N-752, Section -1100 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.

Section -9000 states:

pressure boundary failure: loss of the pressure-retaining function, including leakage, of one or more of the items (e.g., piping, components, tanks) that comprise a piping segment. [emphasis added]

Based on the above, the Staff considers an item to be a component. Therefore, categorization of a subcomponent is not allowed by ASME Code Case N-752.

Note that IWA-9000 of Section XI, ASME Code does define item as a material, part, appurtenance, piping subassembly, component, or component support." However, the definition provides similar examples to components, but does not specify piece-part as is intended to be used in NEI-00-04.

Piece-part in NEI-00-04:

The term piece-part is used in NEI-00-04 in the following paragraphs of NEI-00-04:

Section 4: If the SSCs associated with an ISI segment have already been defined in the risk-informed ISI program, the only additional work is:

43

a. Associate piece parts with a component that has already been categorized in the ISI program and, 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 BPV 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 regarding 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:

44 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. 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 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

45 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.

Enclosure IV also provide the basis for the exemption from the quality assurance requirements in Appendix B.

Based on the above, the differing 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.

Document 2: Memo Establishing DPO Panel

August 12, 2024 MEMORANDUM TO:

Victor Hall, Panel Chair Office of Nuclear Regulatory Research David Roth, Panel Member Office of the General Council Geoffrey K. Ottenberg, Panel Member Region II THRU:

David L. Pelton, Director /RA/

Office of Enforcement FROM:

Gladys Figueroa-Toledo, Differing Views /RA/

Program Manager Office of Enforcement

SUBJECT:

AD HOC REVIEW PANEL - DIFFERING PROFESSIONAL OPINION ASSOCIATED WITH THE NRC AUTHORIZATION OF DUKE ENERGY ALTERNATIVE REQUEST TO USE ASME CODE CASE N-752 AT OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 (DPO-2024-002)

In accordance with Management Directive (MD) 10.159, The NRC Differing Professional Opinion Program; and in my capacity as the Differing Views Program Manager (DVP PM); and in coordination with David Pelton, Director, Office of Enforcement, Andrea Veil, Director, and Office of Nuclear Reactor Regulation; you are appointed as members of a DPO Ad Hoc Review Panel (DPO Panel) to review a DPO submitted by a group of U.S. Nuclear Regulatory Commission (NRC) employees.

The DPO (Enclosure 1) involves the NRC Authorization of Duke Energy Alternative Request to use ASME code case N-752 at Oconee Nuclear Station Units 1, 2, and 3. The DPO has been forwarded to Andrea Veil for consideration and issuance of a DPO Decision.

The DPO Panel plays a critical role in the success of the DPO Program. Your responsibilities for conducting the independent review and documenting your conclusions in a report are addressed in the handbook for MD 10.159 in Section II.F and Section II.G, respectively. The DPO Web site also includes helpful information, such as a Differing Views Best Practices Guide, tables with status information and timeliness goals for open DPO cases, and closed DPO case files (which include previous DPO panel reports). We will also send you additional information that should help you implement the DPO process.

CONTACT:

Gladys Figueroa-Toledo, OE (301) 287-9497

V. Hall, et.al 2

Timeliness is an important DPO Program objective. Thus, the disposition of this DPO should be considered an important and time sensitive activity. Because there is a prior related DPO that has similar concerns raised in this DPO, it is in the best interest of the agency to assign the same panel, to the extent possible. This approach will leverage prior work and expertise and support timeliness, effectiveness, and efficiency. In addition, we are requiring the panel members interview the submitters proposed panel members. Although MD 10.159 identifies a timeliness goal of 75 calendar days for the DPO panel review and report and 21 additional calendar days for the issuance of a DPO Decision, the DPO Program also sets out to ensure that issues receive a thorough and independent review. Therefore, the overall timeliness goal will be based on the significance and complexity of the issues, schedule challenges, and the priority of other agency work. Process milestones and timeliness goals specific to this DPO will be discussed and established at a kick-off meeting to be scheduled in the coming weeks.

Communication of expected timelines and status updates are important in the effectiveness and overall satisfaction with the Differing Views Program. If you need an extension beyond the timeliness goal, please send an e-mail to Mr. Pelton, Ms. Veil, the DPO submitters, and DPOPM.Resource@nrc.gov that includes the reason for the extension request and a proposed completion date.

An important aspect of our organizational culture includes maintaining an environment that encourages, supports, and respects differing views. As such, you should exercise discretion and treat this matter appropriately. To preserve privacy, minimize the effect on the work unit, and keep the focus on the issues, you should simply refer to the employees as the DPO submitters. Avoid conversations that could be perceived as hallway talk on the issue and refrain from behaviors that could be perceived as retaliatory or chilling to the DPO submitters or that could potentially create a chilled environment for others. It is appropriate for employees to discuss the details of the DPO with their co-workers as part of the evaluation; however, as with other predecisional processes, employees should not discuss details of the DPO outside the agency. If you have observed inappropriate behaviors, heard allegations of retaliation or harassment, or receive outside inquiries or requests for information, please notify the Office of Enforcement.

On an administrative note, please ensure that all DPO-related activities conducted by staff are charged to Activity Code ZG0007. Managers should report time to their Management/Supervisor Activity Code. Administrative Assistants should report time to their Secretary/Clerical Activity Code.

V. Hall 3

We appreciate your willingness to serve on the DPO Panel and your dedication to completing a thorough and objective review of this DPO. Successful resolution of the issues is important for the NRC and its stakeholders. If you have any questions or concerns, please feel free to contact me. We look forward to receiving your conclusions and recommendations.

Enclosures:

1. DPO-2024-002 Submittal (ML24215A370)

2. Process Milestones and Timeliness Goals (ML24215A369)

V. Hall 4

SUBJECT:

AD HOC REVIEW PANEL - DIFFERING PROFESSIONAL OPINION ASSOCIATED WITH THE NRC AUTHORIZATION OF DUKE ENERGY ALTERNATIVE REQUEST TO USE ASME CODE CASE N-752 AT OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 (DPO-2024-002)

DATE: AUGUST 12, 2024 DISTRIBUTION:

NON-Public AVeil, NRR MKing, NRR MMitchell, NRR JHoncharik, NRR JCollins, NRR TScarbrough, NRR DWidrevitz, NRR SCumblidge, NRR SClark, OGC RTregoning, RES KCoyne, RES DPelton, OE JCai, OE DSolorio, OE GFigueroa-Toledo, OE ADAMS Package: ML24215A368 MEMO: ML24215A371 - ML24215A370 - ML24215A369 OE-011 OFFICE OE: DPO/PM OE: D NAME GFigueroaToledo DSolorio For DPelton DATE 08/09/2024 08/12/2024 OFFICIAL RECORD COPY

Document 3: DPO Panel Report

Differing Professional Opinion (DPO)

NRC Authorization of Duke Energy Alternative Request to Use ASME Code Case N-752 at Oconee Nuclear Station, Units 1, 2, and 3 (DPO-2024-002)

DPO Panel Report Authors: Victor Hall, Geoffrey K. Ottenberg, David E. Roth Revision: March 21, 2025

Executive Summary Safety is our top priority: The panel concluded that Oconee remains safe under the current use of their authorized alternative based on 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 I, supported by the plants existing operational oversight mechanisms, including risk assessments, audits, and operating experience (OpE) programs. The plants regulatory framework provides sufficient confidence that safety-related functions will continue to be met despite the relatively ambiguous flexibility afforded by N-752.

The panel's review of DPO-2024-002 focused on the application of ASME Code Case N-752 at Oconee and the technical concerns raised by the DPO submitters regarding safety classification and treatment of components. The review is not a comprehensive evaluation of N-752 itself; should the NRC elect to perform a broader review, then such review could include rulemaking.

The panels primary objective was to ensure that the specific application of N-752 that prompted the differing professional opinion did not compromise safety margins or deviate from NRC regulations. Nonetheless, the panel identified key technical concerns that should be addressed to ensure consistent safety practices in future applications of N-752:

A. Minimum Standards: The lack of specific codes and standards to justify alternative methods to the NRC, for repair and replacement activities (i.e., established minimum requirements) is less than ideal for providing clarity, enforcement consistency, and ensuring safety.

B. Categorization of LSS components: There is a potential risk of mischaracterization of components, particularly when alternative treatments are applied to redundant systems or passive subcomponents. The categorization method must be rigorously applied to ensure that safety margins are not undermined.

C. Quality Assurance (QA): The QA requirements in 10 CFR Part 50, Appendix B must remain in place, ensuring that LSS components are subject to appropriate oversight during repair/replacement activities. The panel acknowledges that Oconees application will require use of existing Appendix B processes, and this is consistent with DPO-2024-001's resolution.

D. Flexibility in regulatory processes: While N-752 offers flexibility, it must be evaluated carefully to avoid bypassing critical regulatory steps, such as license amendments under 10 CFR 50.69 for recategorizing components.

Key Findings:

Oconee's approval of N-752 does not raise safety concerns within the plant's regulatory context, based on existing risk-informed practices and safety oversight.

The approval process for Oconee was consistent with NRC standard for safety, although the NRC should review future use through rulemaking or regulatory guide review /

conditional acceptance of the Code Case to ensure its appropriate application across other plants.

The NRCs approval of N-752's application at Oconee was based in part on site-specific details provided by Oconee in RAI responses. The approval is not generic of the usage of N-752.

Recommendations:

The NRC should examine the base set of standards identified through the audit process and meetings with the licensee, with industry input, as part of any future generic review to ensure that minimum safety standards are adhered to.

The categorization process for LSS components should be more structured to account for potential common cause failures and passive-active interactions in redundant systems.

The NRC should consider monitoring the industrys application of N-752s application to evaluate its practical impacts and ensure ongoing safety as its use expands. This could be accomplished through inspection, generic communications, or voluntary initiatives.

The NRC staff should assure that the staff is meeting the Commissions expectations regarding usage of specific regulations when such regulations are available.1 Introduction This report addresses DPO-2024-002, regarding the use of ASME Code Case2 N-752 at Oconee Nuclear Station (ONS). The purpose of this review is to evaluate the technical concerns raised regarding the application of N-752 to ONS, specifically how it impacts safety classifications and treatment of components at the plant. The review is not a comprehensive evaluation of N-752 itself, which requires broader NRC processes, including rulemaking for evaluating the generic acceptability of the code case.

Code Case N-752 allows for alternative treatment of Low Safety Significant (LSS) components, offering some flexibility via categorization and regulatory alternatives. However, the application of N-752 at ONS raises questions about the impact on safety margins and the treatment of passive vs. active components in the plants systems. These issues are critical to ensuring the plants continued safe operation, particularly concerning quality assurance (QA) and repair/replacement requirements under existing NRC regulations.

1 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4) (stating, On a related point, the relationship between the general exemption criteria in § 50.12(a) and other provisions in Part 50 that contain specific exemption criteria or alternative methods of compliance, the Commission would emphasize that § 50.12(a) is the exemption provision that applies generally to the provisions of 10 CFR Part 50. If another regulation in Part 50 provides for specific exemption relief, or for alternative methods of compliance, the criteria of the specific regulation are the appropriate considerations.

(emphasis in original)).

2 The American Society of Mechanical Engineers (ASME) describes the role of code cases thusly: In the event of an urgent need for alternative rules concerning materials, construction, or in-service inspection activities not covered by existing Boiler and Pressure Vessel Code (BPVC) rules, or need for early implementation of an approved code revision, ASME may issue a code case. Code cases are effective immediately upon ASME approval and do not expire. https://www.asme.org/codes-standards/publications-information/code-cases last visited March 6, 2025.

The panels role in this review was to evaluate whether the concerns raised by the DPO submitters are valid in the context of ONS and the specific application of N-752 at the site. The DPO submitters, as leading experts in the field, raised important technical points that are valid and worthy of consideration. The panel carefully considered their invaluable knowledge of code applications and their potential impacts on plant safety.

A tremendous amount of time and effort went into the panel's review. The panel met with the DPO submitters regularly to ensure a thorough understanding of their concerns. Additionally, the panel engaged Senior Level Advisors and met with the approvers of the code alternative to gain insight into their decision-making process. The panel also met with selected inspection and risk-assessment staff to gain their perspectives regarding enforceability of the provisions of the code alternative. The panel also considered the balance with the overall regulatory framework, risk assessments, NRC oversight, operating experience, and outcomes from interactions with various stakeholders, including public meetings and audits.

Background

Overview of ASME Code Case N-752 ASME Code Case N-752 provides an alternative method for categorizing safety-related components in nuclear power plants, allowing components to be classified as Low Safety Significant (LSS) or High Safety Significant (HSS) based on risk-informed principles rather than traditional deterministic methods. The N-752 categorization method evaluates the risk of a pressure boundary component using a consequence analysis that takes as an input a pressure boundary failure of the item being evaluated and determines the effects of that leak or pipe break. It does not evaluate the impact that the alternate treatments to be applied, if the item is ultimately designated as LSS, will have on the active functions or reliability of the associated system. The primary purpose of Code Case N-752 is to offer flexibility in the regulatory framework, particularly in reducing the regulatory burden for LSS components while still ensuring plant safety. However, the application of N-752 raises concerns regarding the adequacy of defense-in-depth, the treatment of passive vs. active components, and potential gaps in quality assurance for LSS systems.

Regulatory Framework The regulatory framework for categorizing and treatment of components and systems in nuclear power plants is governed by several key provisions within the Code of Federal Regulations (CFR):

10 CFR 50.55a: This section outlines the requirements for the application of ASME codes and standards in the construction, inservice inspection, and operation and maintenance of nuclear power plants. It specifies that licensees must adhere to the codes and standards as approved by the NRC. As described in Final Rule, Approval of American Society of Mechanical Engineers Code Cases, 87 Fed. Reg. 11,934, 11,935-6 (March 3, 2022), The ASME develops and publishes the ASME BPV Code that contains requirements for the design, construction, and in-service inspection examination of nuclear power plant components, and the ASME OM Code that contains requirements for inservice testing of nuclear power plant components. The ASME develops code cases that provide voluntary alternatives to BPV and OM Code requirements under special circumstances. The NRC approves the ASME BPV and OM Codes in §50.55a,

Codes and standards via incorporation by reference. Each provision of the ASME Codes incorporated by reference into, and mandated by, §50.55a constitutes a legally binding NRC requirement imposed by rule. The ASME code cases, for the most part, represent alternative approaches for complying with provisions of the ASME BPV and OM Codes. Accordingly, the NRC periodically amends §50.55a to incorporate by reference the NRC's RGs listing approved ASME code cases that may be used as voluntary alternatives to the BPV and OM Codes.

NRCs permissible usage of code cases is controlled via specific licenses and regulations, including 10 CFR 50.55a(a)(3) (saying in part that the use of code cases listed in the NRC regulatory guides in 10 CFR 50.55a (a)(3)(i) through (iii) is acceptable with the specified conditions in those guides when implementing the editions and addenda of the ASME BPV Code and ASME OM Code incorporated by reference in 10 CFR 50.55a(a)(1).), Per 10 CFR 50.55a(a)(3)(ii) (2025), NRC Regulatory Guide 1.147, Revision 21, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, issued March 2024, which lists ASME Code Cases that the NRC has approved in accordance with the requirements in 10 CFR 50.55a(b)(5). Prior revisions of RG 1.147 were listed in prior versions of 10 CFR 50.55a(a)(3)(ii).

Pursuant to 10 CFR 50.55a(z) (alternatives to codes and standards requirements),

alternatives to the requirements of 10 CFR 50.55a(b) through (h) or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. Further, a proposed alternative must be submitted and authorized prior to implementation. 10 CFR 50.55a(z)(1) requires that the applicant or licensee must demonstrate that the proposed alternative would provide an acceptable level of quality and safety; 10 CFR 50.69: This regulation allows for the risk-informed categorization of safety-related components, providing a pathway to classify components as either HSS or LSS based on a comprehensive, multi-pronged review, including use of probabilistic risk assessment (PRA). Under this framework, LSS components designated as Risk-Informed Safety Class (RISC)-3 and RISC-4 may receive relaxed regulatory treatment compared to HSS components, provided that the risk assessments demonstrate that the components do not pose a significant risk to plant safety.

10 CFR 50, Appendix B: This appendix sets forth the quality assurance (QA) requirements for the design, fabrication, construction, testing, procurement, and operation of systems, structures, and components important to safety. While Appendix B requirements still apply, components categorized as LSS under N-752 may not be subject to the same QA standards as those designated as HSS, which raises concerns about the potential reduction in oversight and inspection rigor. The QA requirements in 10 CFR 50, Appendix B are voluntary for licensees that have received approval to implement 10 CFR 50.69.

Historical Precedent The use of risk-informed categorization methods has been an evolving area within nuclear regulatory practices. In 2004, the NRC made effective its final rule in 10 CFR 50.69 to provide 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. The 50.69 rule revised requirements

with respect to special treatment, that is, those requirements that provide increased assurance (beyond normal industrial practices) that SSCs perform their design basis functions. The rule permits licensees (and applicants for licenses) to remove SSCs of low safety significance from the scope of certain identified special treatment requirements and revised requirements for SSCs of greater safety significance.3 The 50.69 rule requires that implementation be performed for an entire system or structure and not for selected components within a system or structure.4 This required scope ensures that all safety functions associated with a system or structure are properly identified and evaluated when determining the safety significance of individual components within a system or structure and that the entire set of components that comprise a system or structure are considered and addressed.5 Per 10 CFR 50.69(b)(2)-(3), usage of the 50.69 process requires prior approval via a license amendment. In the 2004 rulemaking, the Commission explained its decision to use the license amendment process.6 A rulemaking comment suggested that, in light of the desire to move to a more performance-based regulatory regime, voluntary implementation of 10 CFR 50.69 should be developed by licensees using the requirements in the rule and any attendant regulatory guidance, with routine NRC inspection serving to verify acceptable compliance.7 However, the Commission continued to conclude that the review of the license amendment submittal will involve substantial engineering judgment on the part of NRC reviewers, inasmuch as the rule does not contain objective, nondiscretionary criteria for assessing the adequacy of the PRA process, PRA review results and sensitivity studies.8 Accordingly, the Commission required that, consistent with a prior Commission decision, 10 CFR 50.69 requires NRC approval to be provided by issuance of a license amendment.9 The provisions of the risk categorization process in N-752, paragraph I-3.1, include that, The failure consequence can be quantified using the available PRA. It may be considered that the need for substantial NRC reviewer engineering judgement is similarly involved in the review of N-752 Code alternatives for the same reasons as those explained in the 50.69 SOCs.

The 2004 rulemaking discussed how 50.69 relates to 10 CFR Part 50, Appendix B Quality Assurance Requirements.10 Specifically, the rule removes RISC-3 and RISC-4 SSCs from the scope of requirements in Appendix B to 10 CFR part 50.11 Appendix B contains requirements for a quality assurance program meeting specified attributes.12 The intent of Appendix B to 10 CFR part 50, and the complementary regulations, is to provide quality assurance requirements for the design, construction, and operation of nuclear power plants.13 The quality assurance 3 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, 69 Fed. Reg. 68008 (Nov. 22, 2004) (available at https://www.federalregister.gov/d/04-25665/p-3).

4 69 Fed. Reg. at 68,019.

5 69 Fed. Reg. at 68,019.

6 69 Fed. Reg. at 68,015-16.

7 69 Fed. Reg. at 68,015.

8 69 Fed. Reg. at 68,015-16.

9 69 Fed. Reg. at 68,016.

10 69 Fed. Reg. at 68,026.

11 69 Fed. Reg. at 68,026.

12 69 Fed. Reg. at 68,026.

13 69 Fed. Reg. at 68,026.

requirements of Appendix B are to provide adequate confidence that an SSC will perform satisfactorily in service.14 These requirements were developed to be applied to safety-related SSCs.15 In the implementation of Appendix B, a licensee is bound to detailed and prescriptive quality requirements to apply to activities affecting those SSCs.16 As such, these requirements meet the Commissions definition of special treatment requirements.17 Under the 50.69 process, these special treatment requirements are removed from application to RISC-3 and RISC-4 SSCs because the low individual safety significance of those SSCs does not warrant the level of quality requirements that exist in Appendix B.18 In effect, approval of a 50.69 license amendment request makes voluntary many of the special treatment regulatory requirements including 10 CFR 50, Appendix B, while maintaining safety.

The submitters indicated that 10 CFR 50.55a(z) has historically been applied to specific items and specific individual provisions of ASME Code as opposed to a broad class of LSS items and the broad repair/replacement requirements. However, the use of Code Case N-752 introduces some differences in scope and how components are categorized, particularly the potential relaxation of defense-in-depth requirements and the relaxion of certain QA standards for LSS components.

In an earlier DPO, DPO-2024-001, submitters raised similar concerns about the impact of relaxed regulatory requirements on safety, particularly around the categorization of components and the potential for missed failure correlations between passive and active components. The resolution of that DPO highlighted the need for clear, generic communications to clarify that N-752 does not exempt components from Appendix B requirements, ensuring consistent regulatory guidance when adopting new methodologies. The concerns raised in DPO-2024-002 about N-752 echo those earlier points but are specific to its application at Oconee and the potential impacts on component reliability and safety margins.

Panel Evaluation Concern A: Evaluation of minimum standards and recommendations for improvement.

DPO Submitters Summary of Concern:

The alternative treatment in ASME Code Case N-752 to be used in the Oconee Nuclear Station (ONS) alternative request allows the licensee to use any code or standard (whether appropriate or not) or any method. The authorization of N-752 gives the licensee the ability to modify/delete any provisions of that standard or use parts of various standards to develop an owners method for repair and replacement of all safety-related Class 2 and 3 SSCs categorized as low safety significant (LSS). The licensee is also specifically allowed to reduce the NDE requirements to any option, including just plant walkdowns, and removes any Code pressure testing requirements regardless of what standard is used. It does this with no additional technical basis required through their processes and procedures, but simply by categorizing an item as LSS through the N-752 categorization process. In order for the NRC to confirm the level of quality and safety of the licensees proposed alternative, the NRC needs to establish minimum requirements for licensees to apply specific codes and standards 14 69 Fed. Reg. at 68,026.

15 69 Fed. Reg. at 68,026.

16 69 Fed. Reg. at 68,026.

17 69 Fed. Reg. at 68,026.

18 69 Fed. Reg. at 68,026.

or to justify alternative methods to the NRC, for repair/replacement activities during the implementation of Code Case N-752.

Overview The alternative treatment proposed in ASME Code Case N-752 for Oconee Nuclear Station (ONS) provides the licensee with substantial flexibility in selecting codes, standards, or specifications applicable to the items and methods for the repair and replacement of safety-related Class 2 and 3 systems categorized as LSS. This includes the ability to apply the technical requirements of an applicable code or standard, or specification or the use of an applicable post-construction code or standard. The 10 CFR 50.55a(z) alternative to the existing Section XI requirements also permits the Owner to define the Nondestructive Examination (NDE) requirements and pressure testing requirements, potentially lowering the level of scrutiny applied to critical systems when compared to current requirements of fully complying with ASME Section XI. The allowances of 10 CFR 50.69 are similar, since the repair and replacement requirements for ASME Class 2 and Class 3 LSS SSCs in 10 CFR 50.55a(g) are voluntary, when approved to implement 50.69.

While N-752 provides flexibility, it allows these alterations if an evaluation ensuring the structural integrity and leak tightness of the system are sufficient to support the design basis functional requirements of the system is completed. This raised concerns about whether the changes could lead to reduced safety margins for components that are integral to the plants operations since the scope of the required evaluation could be interpreted to be limited to review of the pressure retaining function. The Commission, in the 10 CFR 50.69 SOCs (69 FR 68019),

acknowledged that collectively, RISC-3 SSCs can be safety significant and as such, it is important to maintain their design basis functional capability. Maintenance of RISC-3 design basis functionality is important to ensure that defense-in-depth and safety margins are maintained.

Panel Review and Conclusion for Oconee The panel believes that the NRCs conclusion to approve N-752 for Oconee was justified considering the overall regulatory picture. The decision was based on a comprehensive review that included audits, public meetings, and the knowledge that the plant still will have ongoing oversight, corrective action, and Operating Experience (OpE) programs. These measures provided a foundation for confidence that the plant's safety margins would be maintained. While not perfect, these processes ensure that any issues arising from repairs and replacements can be addressed through corrective actions, with feedback loops built into N-752 requirements.

The NRC staffs conclusion was also informed by an integrated approach that included the continued application of 10 CFR 50, Appendix B requirements, which ensures quality assurance for repair and replacement activities. Oconees RAI responses noted that controls would be established to ensure continued capability and reliability of all design basis functions (including active) potentially affected by a change resulting from a repair/replacement activity using the authorized alternative. The risk categorization method specified by N-752 was also adhered to, helping to ensure that critical pressure boundary safety functions are not compromised.

Need for NRC Review and Minimum Requirements for Generic Approval

While the staffs approval for Oconee was justified based on the plants specific context, generic approval of N-752 should have more specific guidelines. The DPO stems from the uncertainty and a lack of clarity on NRC-endorsed minimum requirements for licensees when selecting codes or justifying alternative methods, particularly when safety-related components are involved. Without this level of oversight, there is a risk that plants could apply unreviewed codes or standards that may not meet the necessary safety standards required of a Code alternative authorized via 10 CFR 50.55a(z)(1), resulting in inconsistencies across the industry.

Regulatory Guide 1.174 presents a four-element approach to evaluating proposed licensing basis changes. With regards to identification of specific codes or standards to be applied as part of the code alternative, Element 3: Define Implementation and Monitoring Program, explains its intent is to ensure that no unexpected safety degradation occurs due to the change to the licensing basis, and that the principal concern is the possibility that the aggregate impact of changes that affect a large class of SSCs could lead to an unacceptable increase in the number of failures from unanticipated degradation, including possible increases in common-cause mechanisms. One member of the panel remains concerned that the statements that the alternative treatments allowed under N-752 can only alter the likelihood of a pressure boundary failure, and not contribute to the increase in a common mode failure of an active function, could distract the implementing licensees change evaluation efforts away from potentially risk-significant adverse effects of the changes, given the wide range of potential codes, standards, or specifications that may be used in repair/replacement activities under N-752.

Further, the burden for review for adequacy of Codes or standards to be applied as a Code alternative is typically borne by the appropriate NRR technical staff, and not inspection staff. The guidance contained in section 4.2 of LIC-102, Review of Relief Requests, Proposed Alternatives, and Requests to Use Later Code Editions and Addenda, indicates that when authorizing the use of Code Cases as alternatives under 10 CFR 50.55a, future unreviewed revisions of the Code Case cannot be authorized without prior NRC review and approval. One panel member believes the intent of this guidance is to ensure that the technical content of any proposed alternative has been reviewed and authorized by the NRC prior to its implementation.

Because Code Case N-752 allows use of unreviewed Codes, standards, or specifications, this will likely result in inspectors questioning the adequacy of the licensees activities performed under the granted alternative. NRC management should be made aware this is likely to result in an inefficient way of arriving at enforcement decisions and may result in an increase in technical assistance requests (TARs) to NRR staff from regional inspection staff. It may also result in inspectors inconsistently evaluating the adequacy of the codes or standards selected for use under the Code Case (e.g., one inspector raising issues on the same code previously evaluated by another inspector without issues) leading to an impression of inconsistent regulation.

Conclusion In the case of Oconee, the NRCs staff conclusion to approve the use of N-752 was justified by the broader regulatory framework, including audits, management oversight, and corrective action programs that provide a solid basis for ensuring safety. However, for potential incorporation into RG 1.147, further scrutiny is necessary to establish clearer guidelines for its application, ensuring consistent safety standards across all plants Concern B: Analysis of risk categorization inadequacies and proposed solutions.

DPO Submitters Summary of Concern:

The categorization process in ASME Code Case N-752 used in the ONS alternative request:

Can designate a component or piece/item of a component (subcomponent) as LSS per EPRI report on ANO valve bonnet, although the Commission-approved 50.69 categorization methodology would categorize the overall component as HSS per 10 CFR 50.69 FRN (69 FR 68019, 68039) including the following risk-informed documents:

o NRC response to public comment C-6 to 50.69 rulemaking (ML042990011) o Section 10.1 vs. Section 10.2 of NEI 00-04 o comment 21 (page 11) to 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)

Takes credit for redundant components/systems that have alternative treatments applied, with a limited basis for failure probability of the alternative treatment items.

This may cause the future N-752 categorization analysis to non-conservatively assess alternative treatment items. Further, the user has the ability to change all redundant systems to LSS with reduced treatment using N 752 in a piece-meal way.

Does not evaluate the effects of the passive subcomponent on the active function of the whole component. The risk informed analysis used to justify not including this reliability/availability evaluation does not itself evaluate the impact failure rate appropriately.

The NRC needs to ensure that the categorization process adequately assigns SSCs (and subparts) to their appropriate HSS or LSS category when implementing Code Case N-752.

For example, the categorization process in Code Case N-752 needs to exclude like-components for remaining capability in defense-in-depth categorization (i.e., it does not allow credit for any identical, redundant SSCs within the system that are classified as LSS and have alternative treatment applied). In addition, an NEI 00-04 assessment, including Section 10.1, needs to be performed for all active function SSCs that could be affected by the failure of the passive SSC on which a licensee intends to utilize the Code Case N-752 categorization process.

Overview The categorization process in ASME Code Case N-752, as applied in the Oconee alternative request, allows for a passive subcomponent of an HSS active parent component to be categorized as LSS. This approach raises concerns about the potential for mischaracterization of components that contribute to safety functions, particularly when considering redundant systems, interactions between passive and active components, and failure effects.

The categorization methodology used under N-752, uses both PRA and qualitative insights to perform a consequence analysis, but does not explicitly require:

Consideration of identical, redundant components/systems that have alternative treatments applied.

An evaluation of how alternative treatments on passive subcomponents may affect the reliability of the parent active component.

A structured approach to assessing the failure probability and impact of alternative treatment under design basis conditions rather than solely relying on industry operating experience (OpE).

While the scope of applicability for N-752s alternative treatment provisions is generally limited to pressure-retaining items, there is an inherent assumption that treatments applied to in-scope components will not impact non-pressure-retaining functions. Indeed, in their 10/20/2023 RAI response, Oconee stated that Alternate Treatment requirements (including repair/replacement) affect only the frequency of passive component failure. If this assumption holds true, it supports the current categorization methodology, which is essentially the same as the passive categorization process used in 10 CFR 50.69 applications. Although the N-752 categorization method is the similar to that used when evaluating passive components under the 50.69 categorization process, the panel noted that the full 50.69 categorization process is more comprehensive and requires defining design basis functions at a system level, assigning those functions to components that perform or contribute to those functions, and requires special defense in depth considerations of the systems active component/functions that are not required by the N-752 process. Specifically, section 6.1 of NEI 00-04 requires that for each active component/function categorized as LSS, an assessment of the remaining mitigation capability without credit for the component proposed as LSS and without credit for any identical, redundant SSCs within the system that are also classified as low safety significant. In this way, even if the alternative treatment affects the SSC/ function being evaluated under 50.69, it is assured as LSS from a defense in depth perspective.

The panel identified plausible cases where alternative treatments on pressure-retaining items or their welded attachments could affect the active function of the larger system, particularly if these interactions are not fully evaluated and addressed by the implementing licensee.

Examples include:

Gate valve guide rails which are welded to a valve body. These internal components contribute to the active function of a valve and operating experience has shown that undesired geometry (e.g., alignment, dimensions) can adversely affect the proper functioning of the valve. (ref. GL 89-10 and EPRI PPM) ASME Section XI, IWA-4120(c)

includes a requirement for items welded to the pressure boundary to comply with Article IWA-4000, which is addressed via the alternate treatments under N-752. Because the valve guide rail does not contribute directly to the pressure boundary function there is potential to overlook the impact of a change on the valves active function.

Thermal binding of gate valves. Body material can affect the expansion/contraction characteristics during design basis temperature changes and can prevent a valve from successfully opening. (ref. GL 95-07 and EPRI TR-114051) If the valve body material is not treated as a critical characteristic and verified appropriately, the potential to introduce a condition preventing an active open function exists.

End loading on active valves. Material changes involving a different thermal coefficient of expansion in long lengths of piping can increase the end loads on active valves due to increased expansion of the piping. The piping loads acting on valves is a known mechanism which could affect the active open/close function of valves. (ref.

https://valve-world-americas.com/how-piping-stresses-affect-valves/ and API Specification 6D)

If a component with a pressure retaining function is determined to be LSS under N-752, then alternate treatments prescribed by the licensee may be implemented. Some reduction of the reliability of an item with alternate treatment being applied may be expected, although the increase in risk due to the treatments is expected to be small, when implementing 10 CFR 50.69, as stated in section II.1.6.2.2, Consistency With the Categorization Process, and II.1.6.5 Basis for RISC-3 SSC Reliability Used in § 50.69(c)(1)(iv) Evaluation, of the SOCs for 10 CFR 50.69 (69 FR 68013 and 69 FR 68014). The panel noted the SOCs addressed the potential for decrease in reliability caused by alternate treatments by clarifying the treatment needs to be consistent with the categorization:

For example, one industry commenter asserted that sensitivity studies eliminate the need to specifically consider SSC reliability changes that might occur due to treatment changes. Another industry commenter stated that cross-system common cause interactions are rarely modeled in PRAs. Similarly, another industry commenter indicated that degradation mechanisms resulting from treatment processes are typically not considered in PRAs. The treatment practices for plant SSCs must support the capability credited in the categorization process for there to be

reasonable confidence that any increase in risk remains small. Therefore, § 50.69(d)(2) was clarified to explicitly require the treatment of RISC-3 SSCs to be consistent with the categorization process.

And The NRC recognizes that the reliability of RISC-3 SSCs could potentially decrease (RISC-3 SSC failure rates increase) due to the reduction in treatment applied to these SSCs as a result of § 50.69 implementation. This is the reason why the Commission requires in the rule that the licensee demonstrate with reasonable confidence that any potential risk increase due to implementation of the rule will be small. However, the NRC also recognizes that it is difficult a priori to relate specific changes in treatment directly to specific changes in SSC reliability. The rule has been constructed to account for this difficulty. First, the categorization process that a licensee uses must comply with the rules requirements. Second, this categorization process will be reviewed and approved by the NRC before implementation. These steps are to have high confidence that SSCs are appropriately categorized so that RISC-3 SSCs are of low individual safety significance. Third, licensees are required to provide reasonable confidence that any risk increase due to implementation is acceptably small and this assessment must be supported by a supporting technical justification that discusses why the assessment adequately addresses the potential reliability changes for RISC-3 SSCs. This basis may include reliance on the capability of the licensees data collection, feedback, and corrective action, which are also addressed by requirements of the rule. Finally, the rule has been revised to clarify the linkage between treatment and categorization and specifically to ensure that the treatment process is consistent with the categorization process, including the risk sensitivity study (i.e., maintain that any risk increase due to reduced treatment is acceptably small).

The feedback and process adjustments required by both the 50.69 and N-752 processes are expected to adjust for known or identified changes in reliability, however, the panel notes that these adjustments are only likely to be made when in-service failures are identified and cannot in all cases account for differences in normal service conditions and accident conditions. This may only be of concern for items where the accident condition is considerably different than normal or test service conditions, such as fluid temperatures and/or timing of component heat up (e.g., sump isolation valves) or differential pressure across a valve (e.g., a valve closed to isolate a pipe rupture or an injection valve into a depressurized vessel).

Panel Review and Conclusion for Oconee For the Oconee alternative request, the panel noted that NRRs Division of Risk Assessment (DRA) examined the overall risk implications, including the categorization process and the impact of alternative treatments on component reliability, and provided a basis for the NRCs confidence in the safety of Oconees application of N-752.

Evidenced by the failure modes and effects analysis requirements of the risk-informed categorization process, the method presented in Code Case N-752 relies on the alternative treatment provisions of the Code Case being only applicable to the pressure boundary aspects of a component, and that it does not change the treatment of items that perform active functions. Consistent with paragraph -1420 of N-752, Oconees 7/27/22 request for the Code alternative stated that the treatment requirements must provide reasonable confidence that each LSS item remains capable of performing its safety-related functions under design basis conditions. The panel understands that the LSS item in this case would have been

determined to only have a pressure boundary/passive safety-related function. However, Code Case N-752 paragraph -1420 requires that the Owner define requirements (emphasis added) to confirm with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design-basis conditions implying that the LSS items to which the alternative treatments will apply could have more than a single pressure boundary function. The alternative requirement in paragraph -1420(c) specifically limits the evaluation of changes as a result of the repair/replacement to structural integrity and leak tightness, potentially limiting the scope of the expected confirmation of effects on safety functions.

As explained by Oconee in the section titled Duke Energy Response to Draft RAI No. 1, Item 2 from Enclosure 2 of the Audit Plan, of their 10/20/23 RAI response, ASME Code Case N-752 does not change the design (e.g. pressure, temperature requirements) of the component and allows only alternatives to the traditional code solutions. Further, Oconees 10/20/23 RAI response explained that While the Relief Request implementation relaxes ASME Section XI and Quality Program Requirements, it does not alleviate Design Control process requirements.

While paragraph -1420(c) of Code Case N-752 only requires evaluation to ensure the structural integrity and leak tightness of the system are sufficient to support the design bases functional requirements of the system Oconees response indicates the Design Control process will be applied ensuring all bounding technical requirements are met. With Oconees described understanding of the N-752 provisions, there is confidence there is low potential for inadvertent common cause failures to be introduced, and that DRAs confirmatory analysis reasonably accounted for the degree of uncertainty of the degradation potentially introduced through use of N-752.

Need for NRC Review of the Generic Issue For generic approval, the NRC should ensure that the methodology used for assigning SSCs (and subcomponents) to HSS or LSS provides a structured and consistent framework that accounts for:

1. Exclusion of like-components for remaining defense-in-depth capability, preventing multiple identical, redundant SSCs from being categorized as LSS without a structured review beyond that required by -1420(c) to ensure that not only the structural integrity and leak tightness of the system are sufficient to support the design bases functional requirements of the system, but also that common mode failures affecting active safety functions are not introduced.

2. An explicit assessment of the impact of passive subcomponent treatment on active functions, ensuring that such interactions do not degrade safety margins.

Conclusion The panel does not recommend overturning the approval of N-752 for Oconee based on the overall regulatory picture, including the examination of risk implications by DRA. The application at Oconee does not pose significant safety concerns, and the plant remains compliant with the necessary safety standards. However, for generic approval of Code Case N-752, the NRC should consider the risk implications, given that common mode failures of active components, as a result of alternative treatments is not a part of N-752 defense-in-depth process. A thorough review should ensure that the categorization process is consistent, structured, and appropriately

addresses all potential safety risks, particularly passive-active interactions and redundant systems.

The NRC should take steps to ensure that the categorization process adequately assigns SSCs (and subcomponents) to the correct HSS or LSS category, and that NEI 00-04 assessments are applied where relevant.

Concern C: Examination of QA compliance issues and necessary clarifications.

DPO Submitters Summary of Concern:

Based on statements in the ANO submittals referenced by Duke Energy that 10 CFR Part 50, Appendix B, does not apply when implementing ASME Code Case N-752, the SE for ONS does not provide assurance that Duke Energy will be legally required to comply with the Quality Assurance (QA) requirements in 10 CFR Part 50, Appendix B, when performing repair and replacement activities for safety-related LSS Class 2 and 3 items when implementing N-752 at ONS. The NRC needs to make clear that all items classified as LSS under N-752 remain within the scope of 10 CFR Part 50, Appendix B, with respect to their design, licensing basis, and regulatory enforcement. The NRC needs to ensure that the licensee shall apply QA activities that provide reasonable assurance that each LSS item within the scope of Code Case N-752 remains capable of performing its safety-related functions under design-basis conditions in accordance with the Appendix B requirements specified by NRR in accepting the ANO 10 CFR 50.54(a)(4) request.

Panel Review The Directors Decision for DPO-2024-001 (ML24311A153) provides key insight into the resolution of QA concerns in the context of alternative treatments. In this earlier case, the NRC clarified that the authorization granted to Entergy to reduce QAPD commitments and use N-752 as a Code Alternative did not relieve the licensee of the responsibility to comply with certain special treatments for Class 2 and 3 passive components. The approval specifically stated that alternative treatments could only be used for demonstrating compliance and must be implemented utilizing existing QA processes and procedures.

Effectively, the alternative treatments granted to Entergy, as part of their QAPD commitment reduction, were approved under a QA alternative that is applicable to Oconee via 10 CFR 50.54(a)(3). The same principle applies to Oconee: while N-752 permits alternative treatment, the plant is still required to comply with Appendix B in the application of these treatments. This ensures that the LSS components classified under N-752 are still subject to the same QA processes to confirm their reliability and safety. In their 10/20/23 RAI response, Duke reiterated that while N-752 allows for flexibility in the QA requirements, other processes such as design control, would remain in place while implementing the code alternative. Further, the staffs 12/13/23 authorization of Oconees proposed alternative explicitly stated that the proposed alternative does not exempt the LSS components from Appendix B requirements.

Conclusion The QA concerns raised in DPO-2024-002 should be addressed through the resolution for DPO-2024-001. The Oconee authorization should similarly ensure that all LSS components under N-752 remain subject to Appendix B requirements, guaranteeing that the licensees QA activities

provide reasonable assurance of the continued performance of these components under design-basis conditions. The panel concurs with the NRCs position that existing QA processes and procedures must be maintained throughout the application of N-752 for Oconee, ensuring the safety and reliability of critical systems.

Concern D: Review of regulatory alignment and implications for NRC practices.

DPO Submitters Summary of Concern:

The request by Duke Energy under 10 CFR 50.55a(z)(1) to use ASME Code Case N-752 at ONS contradicts the Commission's policy, and therefore:

a) The NRC staff cannot authorize the use of 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, the 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, or a separate license amendment in accordance with 10 CFR 50.90.

b) 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. Therefore, 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.

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

Panel Review and Conclusion The Commission expects the intent of its regulations to be met and normally this requires conforming to the regulations as stated.19 The Commission explained that if another regulation in Part 50 provides for specific exemption relief, or for alternative methods of compliance, the criteria of the specific regulation are the appropriate considerations.20 The Commission stated that if the specific exemption criteria, or the alternative methods of compliance, can be satisfied, there is no need also to satisfy the criteria of 10 CFR 50.12(a).21 The Commission recognizes that some Commission regulations are broadly framed and susceptible of various methods of compliance.22 Compliance with the regulations is guided by the use of regulatory guides, branch technical positions, and the standard review plan.23 Such 19 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50765 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

20 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

21 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

22 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

23 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

guidance, however, does not have the force and effect of a regulation.24 The Commission also recognizes that acceptable methods of compliance may change over time based on staff and licensee experience.25 Under the existing regulatory framework, an applicant or licensee may demonstrate that an alternative method of satisfying the regulation is acceptable.26 There are two specific methods in the Commissions regulations that may be used to allow alternatives to certain requirements in specific paragraphs of 10 CFR 50.55a without first obtaining an exemption under 10 CFR 50.12. First, in 10 CFR 50.55a(z), the Commission sets forth a process authorizing alternatives to the requirements of 10 CFR 50.55a(b) through (h) or portions thereof. Second, using the Commissions regulations at 10 CFR 50.69, applicants and licensees may be approved to use 10 CFR 50.69 as an alternative to compliance with the inservice testing requirements in 10 CFR 50.55a(f); the inservice inspection, and repair and replacement (with the exception of fracture toughness), requirements for ASME Class 2 and Class 3 SSCs in 10 CFR 50.55a(g); and the electrical component quality and qualification requirements in Section 4.3 and 4.4 of IEEE 279, and Sections 5.3 and 5.4 of IEEE 603-1991, as incorporated by reference in 10 CFR 50.55a(h).27 Thus a licensee may satisfy the regulations without first obtaining an exemption under 10 CFR 50.12(a) by following the regulations as-written, by using 50.55a(z), or by using 50.69.

If a licensee has been approved for 10 CFR 50.69, then such approval removed certain SSCs from the scope of the controls required under certain paragraphs of 10 CFR 50.55a. Similarly, if a licensee has authorization under 10 CFR 50.55a(z) for an alternative to specific paragraphs of 10 CFR 50.55a, then that alternative also removes the controls otherwise required under the relevant paragraphs of 10 CFR 50.55a. Thus 50.69 and 50.55a(z) both result in a licensee not being required to follow specific paragraphs of 10 CFR 50.55a. Because 50.69 and 50.55a(z) concern alternatives to some of the same requirements in the paragraphs of 50.55a, it is important for licensee requests and NRC approvals and authorizations to be clear and specific.

A licensee with an existing authorization under 50.55a(z) for treatment of certain items can subsequently be approved for 50.69 treatment. In the course of requesting a 50.69 amendment, a licensee could even include a plan for ceasing usage of a 50.55a(z) alternative and thereby put the control fully under 50.69. Once control of an item was under 50.69 rather than the paragraphs listed in 50.55a(z), then a request for an alternative under 50.55a(z) for that item would not be possible because 50.69 would be controlling.

Each method of removing the controls in paragraphs of 10 CFR 50.55a requires prior NRC approval by the NRC, and before making such approval the NRC must first make the requisite safety findings. For a license amendment under 50.69, standards are specified in 10 CFR 50.69(b)(3) and the broader the standards of 10 CFR 50.57 (as applicable under 10 CFR 50.92) including that there is reasonable assurance that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and there is reasonable assurance that such activities will be conducted in compliance with the 24 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

25 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

26 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

27 10 CFR 50.69(b)(1)(v) (available at https://www.ecfr.gov/current/title-10/part-50#p-50.69(b)(1)(v)).

Commissions regulations. For authorization of an alternative under 10 CFR 50.55a(z)(1) the standard is that the proposed alternative would provide an acceptable level of quality and safety in this chapter.

The regulations at 50.55a(z) do not explicitly forbid usage of an alternative that mimics the methods of relief used in other parts of the regulations. However, as previously stated, the Commission has stated that if another regulation in Part 50 provides for specific exemption relief, or for alternative methods of compliance, the criteria of the specific regulation are the appropriate considerations.28 In light of the Commission's statement, the panel compared and contrasted 50.69 with a 50.55a(z) alternative based on ASME Code Case N-752. The panel developed a table of differences to compare the 10 CFR 50.69 process with the provisions of ASME Code Case N-752. This table highlighted the key differences between the two processes, particularly regarding the requirements for recategorizing components, modifying special treatment requirements, and the processes involved in approving code alternatives.

In sum, and as shown on the table, the particular ASME Code Case N-752 50.55a(z) alternative uses methods that similar to the methods used in 10 CFR 50.69. The DPO issue highlighted the similarities and suggested that only a 50.69 process was permissible. The DPO panel agrees that there are similarities and agrees that the similarities raise the question of whether usage of the 50.55a(z) alternative instead of the 50.69 process would meet the Commissions expectation to use the provided regulatory framework when, as the case for 50.69, a framework exists. Thus, to assure that the staff is meeting the Commissions statement that when another regulation in Part 50 provides for specific exemption relief, or for alternative methods of compliance, the criteria of the specific regulation are the appropriate considerations,29 the staff should assure that the Commission is aware of, and approves of, the usage of the ASME Code Case N-752 50.55a(z) alternative.

The panel suggests that the staff should engage the Commission before continuing the practice of utilizing 10 C.F.R. § 50.55a(z)(1) to authorize alternatives that allow a licensee to use processes similar to processes set forth elsewhere in the Commissions regulations. Given the Commissions view that that the regulations that are in-place to provide alternative frameworks are the ones to use to achieve an alternative to a regulation,30 the staff should assure that the Commission is aware of the staffs usage of 50.55a(z)(1) to create an analog to 10 CFR 50.69.

There are several potential paths the staff may wish to use to inform the Commission and to assure that the Commission approves of the staffs continued utilization of 50.55a(z)(1) alternatives as analogues of 10 CFR 50.69. The options below are not intended to be an exhaustive list of paths forward. First, because the staff continues to engage the Commission concerning usage of 10 CFR 50.55a in the context of routine updates and one-offs, the staff could present its desired goal to the Commission. Engaging the Commission on 50.55a(z)(1) alternatives as analogues to other regulations could result in clear direction to the staff and 28 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

29 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

30 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

industry, plus would help to set forth clear criteria for denial or authorization of 50.55a(z)(1) alternatives. The staff could memorialize any Commission direction via updates to the NRCs documents, and industry could similarly use the outcome to update industry guidance that helps industry determine if a 50.69 license amendment request or a 10 CFR 50.55a alternative is the better choice to achieve a licensees goals.

Second, in lieu of utilizing a 10 CFR 50.55a(z)(1) alternative to mimic 10 CFR 50.69, the staff could explore utilizing 10 CFR 50.69 in combination with specific exemptions under 10 C.F.R. § 50.12 that would result in a narrower 50.69 license amendment request (LAR) and, if granted, narrower additional authority. Utilizing a 50.69 LAR with 50.12 exemptions would leverage the substantial guidance available for 50.69 actions. Utilizing a 50.69 with exemptions in lieu of a 10 CFR 50.55a(z)(1) alternative designed to mimic 10 CFR 50.69 would also keep the application and NRC decisionmaking along two well-worn paths. The first path would be the standard license amendment path in 10 CFR 50.90-92 including the associated hearing opportunity, environmental review (typically a categorical exclusion under 10 CFR 51.22), no significant hazards consideration, and standards in 10 CFR 50.69. The second path would be the exemption path wherein the licensee would request to be exempt from the portions of 10 CFR 50.69 it does not seek to adopt under 10 CFR 50.69. The licensee would need to meet 10 CFR 50.12 and, among other things, explain why special circumstances under 50.12(a)(2)(i)-(vii) exist.

The staff has the authority to determine appropriate implementation within the existing regulatory framework. It is important to ensure that this approach aligns with Commission expectations. The DPO panel agrees that the existing regulatory framework allows for the use of 50.55a(z) alternatives for Oconee. The staff already determined that it was appropriate to use 50.55a(z) in this instance, based on its discretion in interpreting the width of the regulation.

However, given the similarity of N-752 to 50.69 processes, it would be prudent for the staff to confirm that the Commission is comfortable with this practice moving forward.

Key Findings and Recommendations Key Findings:

Oconee's approval of N-752 does not raise safety concerns within the plant's regulatory context, based on existing risk-informed practices and safety oversight.

The approval process for Oconee was consistent with NRC standards, although the NRC should review future use through rulemaking or regulatory guide review to ensure its appropriate application across other plants.

N-752's application at Oconee should be seen as an exception, not a precedent for future use, proper regulatory reviews notwithstanding. Generic approval of N-752 will likely require more defined guidelines to ensure consistent safety across various nuclear facilities, particularly in terms of alternative standards used, component categorization, risk and reliability assessments, and adherence to QA standards.

Recommendations:

The NRC should examine the base set of standards identified through the audit process and meetings with the licensee, with industry input, as part of a generic review of N-752, to ensure that safety is maintained.

As part of a generic review of N-752, address the categorization/evaluation processes for LSS components to account for potential cascading failures and passive-active interactions in redundant systems.

The NRC may consider monitoring the industrys application of N-752s application to evaluate its practical impacts and ensure ongoing safety as its use expands. This could be accomplished through inspection, generic communications, or voluntary initiatives.

References Oconee Alternative Request ML22208A031,

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, dated July 27, 2022 ML22235A655,

Subject:

Oconee Nuclear Station, Units 1, 2, and 3 - Acceptance of Requested Licensing Action RE: Proposed Alternative to Use ASME Code Case N-752, dated August 23, 2022 ML23038A183,

Subject:

Oconee Nuclear Station, Units 1, 2, and 3 - Request for Additional Information RE: Alternative Request (RA 0174) to use ASME Code Case N-752, dated February 7, 2023 ML23068A015,

Subject:

Response to Request for Additional Information (RAI)

Regarding Proposed Alternative to Use American Society of Mechanical Engineers Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1, dated March 9, 2023 ML23284A332,

Subject:

Oconee Nuclear Station, Units 1, 2, and 3 - Second Round Request for Additional Information RE: Alternative Request (RA-22-0174) (L-2022-LLR-0060), dated October 11, 2023 ML23293A267,

Subject:

Second Response to Request for Additional Information (RAI)

Regarding Proposed Alternative to Use American Society of Mechanical Engineers Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1, dated October 20, 2023 ML23262A967,

SUBJECT:

OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 - RE:

AUTHORIZATION OF ALTERNATIVE TO USE RR-22-0174, RISK-INFORMED CATEGORIZATION AND TREATMENT FOR REPAIR/REPLACEMENT ACTIVITIES IN CLASS 2 AND 3 SYSTEMS SECTION XI, DIVISION 1 (EPID L-2022-LLR-0060), dated December 13, 2023 ML23278A122, NPHP Oconee Nuclear Station SE Input for RA-22-0174 dated 9/29/2023

Non-concurrence ML23291A227,

Subject:

RE: REQUEST: Concurrence on Oconee N-752 alternative, dated October 18, 2023 ML23331A921, NCP-2023-005 Case File - Public, dated November 16, 2023 ML23331A915, NCP-2023-005 Case File-Non-Public, dated November 16, 2023 N-752 Regulatory Audit ML23178A068,

SUBJECT:

OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 - AUDIT PLAN RE: PROPOSED ALTERNATIVE TO USE ASME CODE CASE N-752, RISK-INFORMED CATEGORIZATION AND TREATMENT FOR REPAIR/REPLACEMENT ACTIVITIES IN CLASS 2 AND 3 SYSTEMS SECTION XI, DIVISION 1 (EPID L-2022-LLR-0060), dated July 3, 2023 ML23269A041, Oconee Nuclear Station, Units 1, 2, and 3 - Code Case N-752 Audit; August 22, 2023, E-mail Providing Additional Information Regarding Design and Quality Program Requirements (EPID L-2022-LLR-0060), dated August 22, 2023 ML23270B836,

Subject:

[External_Sender] Duke Follow-up on Code Explanation and Owners Requirements, dated September 26, 2023 ML23219A140,

SUBJECT:

OCONEE NUCLEAR STATION, UNITS 1, 2 AND 3 - AUDIT REPORT RE: PROPOSED ALTERNATIVE TO USE ASME CODE CASE N-752 (EPID L-2022-LLR-0060), dated October 10, 2023 10 CFR 50.69 References ML12248A035, Vogtle Electric Generating Plant Pilot 10 CFR 50.69 License Amendment Request, dated August 31, 2012 ML14122A364, Vogtle Electric Generating Plant - Unit 1 and Unit 2 Pilot 10 CFR 50.69 License Amendment Request Response to Request for Additional Information, dated May 2, 2014 ML14237A034,

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS RE: USE OF 10 CFR 50.69 (TAC NOS. ME9472 AND ME9473), dated September 17, 2014 NEI 16-09, Risk-Informed Engineering Programs (10 CFR 50.69) Implementation Guidance, Revision 1 NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0

69 FR 68008-68048, Risk-Informed Categorization and reatment of Structures, Systems and Components for Nuclear Power Reactors REGULATORY GUIDE 1.201 (For Trial Use), GUIDELINES FOR CATEGORIZING STRUCTURES, SYSTEMS, AND COMPONENTS IN NUCLEAR POWER PLANTS ACCORDING TO THEIR SAFETY SIGNIFICANCE, dated May 2006 IP 37060, 10 CFR 50.69 RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS INSPECTION, dated October 18, 2022 ASME references ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division I, approval date July 23, 2019 ASME Boiler & Pressure Vessel Code Section XI, 2007 Edition through 2008 addenda, RULES FOR INSERVICE INSPECTION OF NUCLEAR POWER PLANT COMPONENTS Miscellaneous LIC-102, Review of Relief Requests, Proposed Alternatives, and Requests to Use Later Code Editions and Addenda, Revision 3 EPRI 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012 EPRI 3002020763, Consideration of Defense-in-Depth and Safety Margins in Risk Informed Decision Making, August 2021 EPRI TR-110161, Piping System Reliability and Failure Rate Estimation Models for Use in Risk-Informed In-Service Inspection Applications, December 1998 INEL EGG-SSRE-9639, Component External Leakage and Rupture Frequency Estimates, dated November 1991 NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, Revision 3 EPRI Topical Report TR-112657 Revision B-A, Revised Risk Informed Inservice Inspection Procedure. Reference Project #669, dated December 1999 SAFETY EVALUATION REPORT Related to "Revised Risk-Informed Inservice Inspection Evaluation Procedure" (EPRI TR-112657, Rev. B, July 1999), dated October 28, 1999

DRAFT REGULATORY GUIDE DG-1121, GUIDELINES FOR CATEGORIZING STRUCTURES, SYSTEMS, AND COMPONENTS IN NUCLEAR POWER PLANTS ACCORDING TO THEIR SAFETY SIGNIFICANCE, dated May 2003 SECY-23-0061, CLARIFICATION OF THE STAFFS POSITION ON CERTAIN AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE ALTERNATIVES FOR MORE THAN ONE 10-YEAR INSERVICE INSPECTION INTERVAL UNDER TITLE 10 OF THE CODE OF FEDERAL REGULATIONS 50.55a, dated July 21, 2023 Regulatory Guide 1.174, AN APPROACH FOR USING PROBABILISTIC RISK ASSESSMENT IN RISK-INFORMED DECISIONS ON PLANT-SPECIFIC CHANGES TO THE LICENSING BASIS, Revision 3

This table highlights the areas where Code Case N-752 may pose additional safety risks due to its lack of comprehensive system assessments.

Aspect 50.69 Approach Code Case N-752 Safety Concern DPO Issues Scope of Categorization Requires the categorization of entire systems (active and passive components/system functions) as HSS or LSS. Detailed SSC categorization performed by IDP allows for component/subcomponent categorization.

Allows individual items (pipe segments, subcomponents, supports) to be categorized as LSS or HSS without considering reliability effect on active components/functions.

Risk of incomplete safety assessment: Potential miscategorization of subcomponents leading to inadequate safety treatment.

B The passive categorization process requires assuming failure of the passive component and evaluating direct and indirect effects. This ensures that system-level impact is still considered (for loss of pressure boundary failures only),

even if categorization is at the component level.

Treatment of Redundant Systems Does not allow credit for identical, redundant active SSCs within the system that are also classified as LSS in the core damage DID categorization process*NOTE 0.

Credits redundant passive components that are categorized as LSS and unaffected by the direct and indirect effects of a loss of pressure boundary, potentially reducing their safety treatment.

Increased failure risk:

Miscrediting redundancy if common cause failure mechanism or other adverse effect introduced through treatment may weaken defense-in-depth and increase vulnerability to failures.

B, D Redundancy is a valid risk-mitigation measure. NRC-endorsed NEI guidance for 50.69 already allows for similar treatment. Inspection and oversight programs ensure that licensees do not inappropriately classify redundant safety-significant components.

Methodology for Categorization Uses system-based categorization, evaluating qualitative, probabilistic, and passive elements of components of the entire system.

Allows Item-based categorization, evaluating each component or subcomponent individually considering only loss of pressure boundary failure modes and their effects.

Uncertainty in risk assessment: Does not account for system-wide/

common mode failures introduced through alternate B, D

Aspect 50.69 Approach Code Case N-752 Safety Concern DPO Issues treatment, potentially leaving active functions exposed.

50.69 already separates active and passive categorization methodologies *NOTE 1a, 1b, and NRC has found this approach acceptable when categorization is done by an IDP. Passive components are assumed to fail in risk evaluations, addressing concerns about system-wide impacts.

Defense-in-Depth Evaluates defense-in-depth more comprehensively with multiple layers of protection (includes active and passive functions). Considers potential for alternate treatments to be applied to identical, redundant active components when assessing level of defense in depth.

Addresses defense in depth necessary during leakage/rupture events, but does not consider how identical/redundant component treatment affects overall defense.

Reduced safety margins:

Removing layers of defense may leave gaps in safety, leading to unanticipated vulnerabilities.

A, B, C

The categorization process retains key safety functions for high-risk components. Code Case N-752 maintains defense-in-depth for high-risk components, but it optimizes resource allocation for low-safety-significant systems, leveraging risk insights to focus on safety-critical elements. *NOTE 2, 3 Quality Assurance (QA)

Makes 10 CFR 50, Appendix B optional for LSS components, also removing requirements for IST, ISI, and repair/replacement for Class 2 & 3 RISC-3 items.

Requires alternative QA measures for LSS components under Appendix B, ensuring compliance with requirements but relaxing application of QA standards.

Compromised safety:

Relaxation of QA controls increases the chance of unnoticed degradation or failure in critical components.

Introduces uncertainty in level of dedication activities applied to purchased items.

C Appendix B change control processes remain applicable. Licensees must evaluate any changes to configuration, design, materials, fabrication, examination, and pressure-testing requirements, and NRC inspections ensure appropriate implementation. Additionally, NRC has accepted similar risk-informed reductions in special treatment requirements under 50.69, with 50.69 additionally eliminating the Appendix B process requirements.

Aspect 50.69 Approach Code Case N-752 Safety Concern DPO Issues Inservice Testing (IST) and Inservice Inspection (ISI)

Makes IST requirements of 50.55a(f) and ISI requirements for Class 2 and 3 components of 50.55a(g) optional.

Maintains full IST and ISI requirements for LSS systems, ensuring component integrity over time. Applicable ISI requirements may, as a result of use of a new construction code or standard, impose different requirements following the repair or replacement activity.

Insufficient monitoring:

Reduced inspection due to the selected code, standard, or specification for the repair/replacement item could miss signs of degradation.

A, C RI-ISI has already been approved. N-752 still requires periodic inspection and timely corrective actions, which provide reasonable assurance of component integrity, assuming the applicable code or standard selected for the repair/replacement activity provides for examination.

Reliability of Components Assesses the reliability of components based on their functionality within the system as a whole.

Assumes passive functions are the only potentially affected functions due to alternative treatment, potentially ignoring failure correlations.*NOTE 4 Increased likelihood of undetected failures: Passive components could fail without proper detection, impacting system performance.

B, D Passive component failures are assumed in the risk evaluation. The methodology explicitly evaluates direct and indirect impacts on active components, ensuring that safety functions are maintained during loss of pressure boundary failure modes. While N-752 assumes passive component failure, it also examines potential cascading effects on active components to ensure no systemic vulnerabilities are overlooked during loss of pressure boundary failure modes. Due to NDE and pressure testing optionally being defined by Owners requirements, uncertainty in reliability of the pressure boundary components is introduced. The alternate treatments under 50.69 could also introduce the same uncertainty.

Regulatory Relief Grants extensive regulatory relief, allowing flexibility in choosing or modifying codes for repairs, with less oversight.

Allows for QA and ASME Code alternatives for repair/replacement of LSS pressure boundary items, while maintaining Appendix B, IST, and ISI program requirements, among others, and applying nationally recognized codes, standards, or specifications.

Increased potential for substandard repairs:

Flexibility in choosing codes without prior NRC review of such codes introduces uncertainty in ultimate quality and regulatory acceptability.

The alternate treatments under 50.69 could also A, C

Aspect 50.69 Approach Code Case N-752 Safety Concern DPO Issues introduce the same uncertainty.

Changes to codes and standards must still go through engineering justification and change control processes, ensuring the pressure boundary function is not compromised. NRCs oversight via inspections helps ensure proper implementation. While N-752 allows greater flexibility in choosing codes, it still currently requires NRC approval of a Code alternative and oversight for changes that affect safety-critical components. This process mitigates concerns about substandard repairs.

Risk Assessment Requires a comprehensive risk assessment at system level considering active (probabilistic), passive, and qualitative aspects to evaluate the preliminary categorization. Final categorization by IDP can consider an item with a purely passive function LSS even if its system has HSS active functions.

Passive Risk assessment process not substantively different than the passive part of the risk assessment done using 50.69 methods. Based on consequence evaluation assuming loss of pressure boundary. Is reliant on repair/replacement treatment not affecting the active functions or reliability.

Incomplete risk picture:

Focusing only on individual items may fail to account for cascading failures or system-wide risks introduced by alternate treatment. FMEA required by N-752 categorization process limited to pressure boundary considerations.

A, B, D

The same passive categorization methodology was previously accepted in ANO2 and TR 3002025288, showing precedent for approval. Risk is inherently bounded by conservative assumptions, making it unlikely that system-wide risks are overlooked as long as treatment options are prevented from having unaccounted for affect on active functions.

50.69 and Code Case N-752 Comparison Table

• NOTE 0: see NEI 00-04, section 6.1, and Farley NCV 05000348,05000364/2024003-02 (ML24317A088)

• NOTE 1a: The NRC endorsed guidance in NEI 00-04, section 10.2, allows engineering and system analyses to identify specific component level or piece part functions and importance for the safety-significant SSCs. However, the guidance suggests that licensee must determine that There is no credible failure mode for the SSC that would prevent a safety-significant function from being fulfilled. The specific FMEA required by N-752 does not require consideration of failure of the non-pressure boundary functions unless induced by a pressure boundary failure.

• NOTE 1b: Vogtle stated in an RAI response (ML14122A364) related to its 10 CFR 50.69 amendment request, that the final risk of a component having both passive and active functions would be the higher of the active risk, passive risk, or defense in depth.

• NOTE 2: The 10 CFR 50.69 SOCs (69 FR 68019) credits the maintenance of the design basis functions and the 50.69(d)(2) requirement as providing for adequate DID and Safety Margins: Maintenance of RISC-3 design basis functionality is important to ensure that defense-in-depth and safety margins are maintained.

• NOTE 3: Defense in Depth considerations during categorization under 50.69 (IDP considerations in section 9.2.2 of NEI 00-04) and under N-752 (paragraph I-3.4.2(b)(6)) are nearly identical, and both require potential for common cause failures is taken into account in the risk analysis categorization. One distinction is that in the complete NEI 00-04 process, system functions (including active) are mapped to the SSCs being categorized and this is not done under N-752 as only the passive risk is being categorized.

• NOTE 4: Potential for common cause failures is part of the defense in depth review. Evidenced by the DID review done by ANO during their EFW governor valve bonnet replacement activity, the focus of the common cause failure is related to the consequence analysis which postulates only a pressure boundary loss and does not require consideration of any potential adverse affects of the alternate treatments (ref. ANO EC 77577 and CALC-ANO2-EP-21-00002 Revision 0): Common cause is a fundamental aspect of the consequence evaluation methodology and therefore is taken into account.

Mapping of DPO Concerns:

A - Risk-Based vs. Risk-Informed Approach: Focuses on the reliance on risk assessment without adequate defense-in-depth and monitoring strategies.

B - Passive vs. Active Function Interaction: Highlights concerns that subcomponent failures could impact active system functions.

C - Removal of Quality Assurance Requirements: Questions whether removing QA standards ensures an acceptable level of quality and safety.

D - Categorization and Redundant System Considerations: Covers concerns about crediting redundancy, categorizing subcomponents separately, and the potential for misclassification of high safety significant (HSS) components.

Key Takeaways:

1. 50.69 ensures more comprehensive risk evaluation but with less stringent alternative treatment requirements.

2. Code Case N-752 offers greater flexibility than current requirements, but at the cost of reduced defense-in-depth assurance and QA standards potentially leading to safety gaps in crucial components, especially when failures in passive parts can cascade to affect active functions.

3. Precedent Matters: NRC has previously accepted similar methodologies under 10 CFR 50.69, and Code alternative ANO2-R&R-004.

4. Defense-in-Depth is Maintained: Even though N-752 relaxes certain requirements, it still evaluates available redundancy assuming a loss of pressure boundary failure.

5. Risk Assumptions are Conservative: Failure assumptions are built into the categorization process to mitigate concerns about passive component failure effects on other equipment. Reliability of the redundancy is maintained by the N-752 provision to confirm with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design-basis conditions.

6. Regulatory Oversight Exists: NRC maintains authority to challenge misapplications through inspections and engineering change control processes.

Document 4: DPO Decision

MEMORANDUM TO:

Those on the Attached List FROM:

Michael F. King, Acting Director Office of Nuclear Reactor Regulation MICHAEL KING Digitally signed by MICHAEL KING Date: 2025.04.28 07:36:57 -04'00'

SUBJECT:

DIFFERING PROFESSIONAL OPINION DECISION ON THE U.S.

NUCLEAR REGULATORY COMMISSION (NRC)

AUTHORIZATION OF DUKE ENERGY ALTERNATIVE REQUEST TO USE OF AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) CODE CASE N-752 AT OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 (DPO-2024-002)

The purpose of this memorandum is to respond to your differing professional opinion (DPO) submitted on May 7, 2024, in accordance with Management Directive 10.159, The Nuclear Regulatory Commission Differing Professional Opinions Program (Agencywide Documents Access and Management System Accession No. ML23123A099). In your DPO titled, NRC authorization of Duke Energy Alternative Request to use ASME Code Case N-752 at Oconee Nuclear Station, Units 1, 2, and 3 (ML24215A370), concerns were raised regarding the impact of the application of Code Case N-752 on safety classifications and treatment of components at Oconee Nuclear Station, Units 1, 2 and 3 (ONS).

Specifically, you raised concerns that the alternative treatment in ASME Code Case N-752 to be used in the ONS (1) allows the licensee to use any code or standard (whether appropriate or not) or any method, (2) allows for the potential for mischaracterization of components that contribute to safety functions, particularly when considering redundant systems, interactions between passive and active components, and failure effects, and (3) allows the licensee to misinterpret the NRC approval as granting an exemption from the quality assurance (QA) requirements in 10 CFR Part 50, Appendix B, for safety-related Class 2 and Class 3 items categorized as low safety significant (LSS) when implementing Code Case N-752. Additionally, you raised concerns that the NRC approval to implement Code Case N-752 at ONS contradicts the Commissions policy because an approval under the 10 CFR 50.55(z) alternative process bypasses 10 CFR 50.69(b)(2), which requires a license amendment to recategorize structures, systems and components (SSCs) and change special treatment for safety-related SSCs. You contend that these issues raise questions about the impact on safety margins and the treatment CONTACT: Jordan Hoellman, NRR 301-415-5481 April 28, 2025

Those on the Attached List of passive versus active components in the plants systems and are critical to ensuring the plants continued safe operation, particularly concerning QA and repair/replacement requirements under existing NRC regulations.

I found your DPO to be carefully researched and of extremely high quality. I commend you for your commitment and dedication to the NRCs mission. Your willingness to raise concerns with your colleagues and managers and ensure that your concerns are heard and understood is admirable and vital to ensuring a healthy safety culture within the agency.

My response to the DPO is described in the enclosure and was informed by the DPO Panel report and their recommendations (ML25083A031).

Enclosure:

Directors Decision for Differing Professional Opinion

MEMORANDUM TO THOSE ON THE ATTACHED LIST Office of Nuclear Reactor Regulation:

Matthew A. Mitchell, Branch Chief Piping and Head Penetrations Branch Division of New and Renewed Licenses John Honcharik, Senior Materials Engineer Piping and Head Penetrations Branch Division of New and Renewed Licenses Jay Collins, Senior Materials Engineer Piping and Head Penetrations Branch Division of New and Renewed Licenses Dan Widrevitz, Materials Engineer Vessels and Internals Branch Division of New and Renewed Licenses Stephen E. Cumblidge, Materials Engineer Piping and Head Penetrations Branch Division of New and Renewed Licenses Thomas G. Scarbrough, Senior Mechanical Engineer Mechanical Engineering & Inservice Testing Branch Division of Engineering and External Hazards

Package: ML25114A104 Memorandum: ML25114A108 Director's Decision: ML25114A109

• via email NRC-002 OFFICE NRR*

NAME MKing DATE 04/28/2025

DIRECTORS DECISION FOR DIFFERING PROFESSIONAL OPINION ON THE U.S.

NUCLEAR REGULATORY COMMISSION (NRC) AUTHORIZATION OF DUKE ENERGY ALTERNATIVE REQUEST TO USE OF AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) CODE CASE N-752 AT OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 (DPO-2024-002)

Background

In DPO-2024-002, titled, NRC authorization of Duke Energy Alternative Request to Use of ASME Code Case N-752 at Oconee Nuclear Station, Units 1, 2, and 3 (Agencywide Documents Access and Management System Accession No. ML24215A370), concerns were raised regarding the impact of the application of Code Case N-752 on safety classifications and treatment of components at Oconee Nuclear Station, Units 1, 2 and 3 (ONS). Specifically, you raised concerns that the alternative treatment in ASME Code Case N-752 to be used in the ONS (1) allows the licensee to use any code or standard (whether appropriate or not) or any method, (2) allows for the potential for mischaracterization of components that contribute to safety functions, particularly when considering redundant systems, interactions between passive and active components, and failure effects, and (3) allows the licensee to misinterpret the NRC approval as granting an exemption from the quality assurance (QA) requirements in 10 CFR Part 50, Appendix B, for safety-related Class 2 and Class 3 items categorized as low safety significant (LSS) when implementing Code Case N-752. Additionally, you raised concerns that the NRC approval to implement Code Case N-752 at ONS contradicts the Commissions policy because an approval under the 10 CFR 50.55(z) alternative process bypasses 10 CFR 50.69(b)(2), which requires a license amendment to recategorize structures, systems and components (SSCs) and change special treatment for safety-related SSCs. You contend that these issues raise questions about the impact on safety margins and the treatment of passive versus active components in the plants systems and are critical to ensuring the plants continued safe operation, particularly concerning QA and repair/replacement requirements under existing NRC regulations.

The DPO Ad Hoc Review Panel (the Panel) issued their report to me on March 21, 2025, after reviewing the applicable documents, conducting internal interviews with relevant individuals, and completing their deliberations (ML25083A031).

To inform my decision regarding this DPO, I reviewed the DPO submittal, the Panels report, and applicable documents. I also considered the discussions we had with the DPO submitters and relevant individuals during the Rapid Resolution phase of the DPO process. Further, I discussed specific aspects of the Panels recommendations with appropriate subject matter experts.

Summary of Issues The Panel focused on the application of ASME Code Case N-752 at ONS and the technical concerns raised by the DPO submitters regarding safety classification and treatment of components. The Panels primary objective was to ensure that the application of Code Case N-752 does not compromise safety margins or deviate from NRC regulations. The Panel identified three key findings and four recommendations. The Panels evaluation identified four concerns agreed upon by the DPO submitters. My assessment of the Panel conclusions will follow the format of the Panels report.

My Assessment of the Panel Conclusions The Panel performed a thorough review of the DPO and related technical areas. Their report focused on safety as the top priority and relied on key insights from the resolution of DPO-2024-001 to address the QA concerns raised in this DPO. The report was well written and provided key findings and recommendations from the Panel. I appreciate the Panels thoughtful assessment of the concerns raised in the DPO.

Before detailing my decisions related to the specific conclusions in the Panels report, it is important to address the risk-and safety-significance aspects of the issues raised. This is important given Commissions direction in SRM-SECY-19-0036, application of the single failure criterion to NuScale actuation block valves, which directs the staff to 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 as part of any licensing review or other regulatory decision. The decision is also informed by the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy (ADVANCE) Act legislation that requires the agency to assess potential updates to the Differing Professional Views or Opinions process to ensure any impacts on agency decisions and schedules are commensurate with the safety significance of the differing opinion.

Code Case N-752 provides for risk-informed categorization of pressure retaining functions (i.e., passive function) of components as high safety significant (HSS) or LSS consistent with the guidance in Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (ML17317A256). Code Case N-752 requires the Owner to define alternative treatment requirements, which will confirm with reasonable confidence that each LSS item remains capable of performing its safety-related function, including the passive function of components that have both active and passive functions (e.g., valves and pumps). Code Case N-752 does not change treatment on the active functions of components.

During the internal discussions of the ONS Code Case N-752 review, NRC staff in the Division of Risk Assessment (DRA) performed an independent Standardized Plant Analysis Risk (SPAR) evaluation to develop a conservative upper bound to quantify the risk of the staff concerns over application of Code Case N-752. Multiple cases were developed and modeled to quantify the impact to internal events core damage frequency (CDF), including an across the board threefold, fivefold, and tenfold increase in failure rate frequency to account for uncertainty in relaxed treatment requirements. The staff included Class 1, 2, and 3 pressure retaining SSCs in its bounding assessment. Although Code Case N-752 only includes Class 2 and 3 pressure retaining SSCs, DRA conservatively included Class 1 pressure retaining SSCs; and, even with this conservatism, the analysis showed that the increase in CDF for each of those increased failure frequencies fell below the RG 1.174 change threshold of 1E-5/yr.

Although failure rates of passive components are typically three to four orders of magnitude lower than active components, the methodology assumes 100% failure of the passive component and assesses categorization conservatively based on consequence alone. Furthermore, the difference between active and passive components is greater for active common cause failure rates because the likelihood of multiple passive components failing simultaneously is very low. The consequence evaluation addresses both direct and indirect impacts to the plant, including upstream and downstream components, as well as surrounding components outside the system that are affected by spatial effects, such as pipe whip or flooding. The sum total of the resulting failures is analyzed against quantitative indices as well as deterministic considerations. Since alternative treatment requirements, among other stipulations, must confirm that each LSS item remains capable of performing its safety-related function under design basis conditions, the failure rates of passive components are not expected to drastically increase and challenge the LSS categorization determination. It should also be noted that Code Case N-752 provides an additional layer of defense in depth compared to the handling of LSS components for licensees who voluntarily adopt 10 CFR 50.69. Different than 10 CFR 50.69, it does not provide an exemption from other requirements, including 10 CFR 50, Appendix B. It only approves alternative treatments for demonstrating compliance, does not relieve licensees of certain special treatments for Class 2 and 3 passive components, and requires the implementation of alternative treatment utilizing the existing QA processes and procedures. Therefore, the NRC staff has confirmed that the LSS items within the scope of the DPO concerns have appropriately and adequately been determined to be of low risk-and safety-significance and with defense in depth to address the additional uncertainties associated with implementation of Code Case N-752 compared to 10 CFR 50.69. Further, the Panels report recognizes that ONS remains safe under the current use of their authorized alternative based on ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division I, supported by the plants existing operational oversight mechanisms, including risk assessments, audits, and operating experience (OpE) programs.

However, the Panels assessment highlighted that the flexibility provided by Code Case N-752 allows alternative treatment in selecting codes, standards, or specifications applicable to the items and methods for the repair and replacement of safety-related Class 2 and 3 systems categorized as LSS without requiring additional technical justification or NRC review beyond the categorization of items as LSS. This could result in reduced safety margins for components that are integral to the plants operations since the scope of the required evaluation could be interpreted to be limited to review of the pressure retaining function. The Panels assessment also highlighted that the categorization process in Code Case N-752, as applied in the ONS alternative request, could result in mischaracterization of components that contribute to safety functions. My assessment focuses on addressing these potential concerns.

Concern A: Evaluation of minimum standards and recommendations for improvement.

As expressed by the Panel, this concern relates to the alternative treatment proposed in ASME Code Case N-752 for ONS providing the licensee with substantial flexibility in selecting codes, standards, or specifications applicable to the items and methods for the repair and replacement of safety-related Class 2 and 3 systems categorized as LSS. This flexibility allows these alternatives without requiring additional technical justification or NRC review beyond the categorization of items as LSS, which raises concerns about whether the changes could lead to reduced safety margins for components that are integral to the plants operations since the scope of the required evaluation could be interpreted to be limited to review of the pressure retaining function. In their report, the Panel concluded that the approval of Code Case N-752 for ONS was justified considering the overall regulatory picture. The decision was based on a comprehensive review of related audits and public meetings, and the knowledge that the plant will have ongoing oversight, corrective action, and OpE programs. These measures provided a foundation for confidence that the licensee will maintain sufficient plant safety margins. One member of the panel expressed that they remain concerned that the statements that the

alternative treatments allowed under Code Case N-752 can only alter the likelihood of a pressure boundary failure and not contribute to the increase in a common mode failure of an active function. The panel member noted that this could distract the implementing licensees change evaluation efforts away from potentially risk-significant adverse effects of the changes, given the wide range of potential codes, standards, or specifications that may be used in repair/replacement activities under Code Case N-752. Because Code Case N-752 allows use of unreviewed Codes, standards, or specifications, NRC inspectors may question the adequacy of the licensees activities performed under the granted alternative. This may result in an inefficient way of arriving at enforcement decisions and may result in an increase in technical assistance requests (TARs) to NRR staff from regional inspection staff. It may also result in inspectors inconsistently evaluating the adequacy of the codes or standards selected for use under the Code Case (e.g., one inspector raising issues on the same code previously evaluated by another inspector without issues) leading to an impression of inconsistent regulation. To address this concern, the Panel recommended potential incorporation into RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, (ML19128A244) to provide necessary further scrutiny to establish clearer guidelines for its application, which would endorse the code case and ensure consistent safety standards across all plants. As discussed further in my assessment, while I agree that there are benefits in NRC providing generic approval of the code case, I disagree that additional constraints beyond those already included in the code case are necessary.

I agree with the Panels conclusion that the NRCs approval is justified by the broader regulatory framework, including audits, management oversight, and corrective action programs that provide a solid basis for ensuring safety. I appreciate that the DPO concern stems from the perceived uncertainty and a lack of clarity on NRC-endorsed minimum requirements for licensees when selecting codes or justifying alternative methods, particularly when safety-related components are involved. However, based on the low safety significance of the components addressed by Code Case N-752, the staffs evaluation of the alternative procedures the licensee committed to, and that the licensee must confirm that each LSS item remains capable of performing its safety-related function under design basis conditions, I think it is appropriate to provide flexibilities to licensees to implement Code Case N-752 consistent with the NRCs approval. These flexibilities will allow licensees to complete timely repair/replacement of LSS components to support safe operations, while ensuring that the component remains capable of performing its safety-related function under design basis conditions. It is also important to note that the scope of the approved alternative does not impact the level of QA applied to active features, so I do not agree with the concerns raised that the scope of the required evaluation could be interpreted to be limited to review of the pressure retaining function.

I agree, in part, with the Panel observation that the NRCs inspection and oversight programs could reveal areas of ambiguity regarding how to interpret the requirements for implementing Code Case N-752. However, we have successfully adapted our regulatory guidance and inspection and oversight programs when similar new risk-informed initiatives were adopted, and I am confident we will be able to do the same with this new program. I do see some benefit to providing some additional clarity to inspection staff to ensure we are taking an appropriate risk-informed and performance-based approach to oversight of licensees implementation of this new program. Based on the low safety significance of the components addressed by Code Case N-752, any inspection should be conducted using existing baseline inspection samples (i.e., resource neutral) and should not be prioritized over more significant inspection samples.

Any inspection should focus on the actual performance of any SSCs receiving alternative treatments and associated ongoing required performance monitoring and corrective actions

versus attempting to inspect the appropriateness of codes selected by licensees. I am also confident that appropriate use of the very low safety significance issue resolution (VLSSIR) process will enable the NRC to remain focused on safety significant implementation issues and avoid expending significant resources on very low safety significant implementation questions.

Concern B: Analysis of risk categorization inadequacies and proposed solutions.

As expressed by the Panel, the categorization process in ASME Code Case N-752, as applied in the ONS alternative request, allows for a passive subcomponent of an HSS active parent component to be categorized as LSS. This approach raises concerns about the potential for mischaracterization of components that contribute to safety functions, particularly when considering redundant systems, interactions between passive and active components, and failure effects. In their report, the Panel acknowledged that while the Code Case N-752 categorization method is the similar to that used when evaluating passive components under the 10 CFR 50.69 categorization process, the full 10 CFR 50.69 categorization process is more comprehensive and requires defining design basis functions at a system level, assigning those functions to components that perform or contribute to those functions, and requires special defense in depth considerations of the systems active component/functions that are not required by the Code Case N-752 process. The Panel also proposed some plausible cases where alternative treatments on pressure-retaining items or their welded attachments could affect the active function of the larger system, if these interactions are not fully evaluated and addressed by the implementing licensee. Like in response to Concern A, the Panel concluded that the approval of Code Case N-752 for ONS was justified considering the overall regulatory picture, including the examination of risk implications by DRA. The Panel noted that for generic approval of Code Case N-752, the NRC should consider the risk implications, given that common mode failures of active components, as a result of alternative treatments, is not a part of the Code Case N-752 defense-in-depth process. The Panel recommended that a thorough review should ensure that the categorization process is consistent, structured, and appropriately addresses all potential safety risks, particularly passive-active interactions and redundant systems.

I agree with the Panels conclusion that the approval of Code Case N-752 for ONS was justified considering the overall regulatory picture, including the examination of risk implications by DRA.

I also agree with the Panel that the categorization method of Code Case N-752 is similar to, but deliberately not as comprehensive as the categorization method of 10 CFR 50.69, and an approval to implement Code Case N-752 should not be seen as equivalent to an approval to implement 10 CFR 50.69. The scope of Code Case N-752 is limited to only passive components and provides defense in depth beyond what the NRC found acceptable for similar LSS passive components under 10 CFR 50.69. Specifically, the use of Code Case N-752 maintains 10 CFR 50, Appendix B. The use of 10 CFR 50, Appendix B can play a vital role in preventing common cause failures. This involves meticulous planning, design reviews, and appropriate verification processes to ensure that design changes are robust and less susceptible to failures that could affect multiple systems simultaneously. These measures would in principle encompass prevention of common mode failures where multiple components fail in the same way, often due to a single common cause or a shared environmental condition.

In addition, inservice testing (IST) and inservice inspection (ISI) program requirements, among others, remain for use of N-752. In contrast, 10 CFR 50.69 provides extensive relief from the above requirements for LSS systems. The above measures provided for in the approvals for use of Code Case N-752 beyond what is required under 10 CFR 50.69 help to provide confidence that there is low potential for inadvertent common cause failures of safety significance to be introduced through interactions between passive and active components.

Combined with the licensees requirement to confirm that each LSS item remains capable of performing its safety-related function under design basis conditions provides confidence that components are appropriately characterized. I acknowledge that the Panel provided examples where alternative treatments on pressure-retaining items, or their welded attachments could affect the active function of the larger systems. However, I did not find them compelling arguments for or against the use of Code Case N-752, because each case would have been prevented or mitigated if the licensee was following requirements that would remain in place with or without use of Code Case N-752 (e.g., 10 CFR 50, Appendix B requirements).

While I do see the benefit in NRC providing generic approval of the code case, I disagree that additional constraints beyond those already included in the code case are necessary. I see no evidence that other licensees who choose to request approval to implement Code Case N-752 would depart from the specified requirement to using the QA Programs and procedures that ONS clarified that they would follow. All NRC licensees have similar NRC-inspected corrective action programs and design change control processes that are required to be followed.

Concern C: Examination of QA compliance issues and necessary clarifications.

As expressed by the Panel, the concern relates to the licensees perceived misunderstanding that the NRC approval granted an exemption from the requirements of 10 CFR 50, Appendix B.

In their report, the Panel observed that the resolution of DPO-2024-001 highlighted the need for clear, generic communications to clarify that Code Case N-752 does not exempt components from 10 CFR 50, Appendix B requirements, ensuring consistent regulatory guidance when adopting new methodologies.

I agree with the Panels conclusion that the QA concerns raised in this DPO should be addressed through the resolution for DPO-2024-001. On February 10, 2025, in response to the Directors Decision on DPO-2024-001, the NRC issued NRC Information Notice 2025-01:

Lessons Learned when Implementing ASME Code Case N-752 (ML24323A057), specifically to inform licensees and permit holders of recently observed inconsistencies between the language in licensee programs during the implementation of Code Case N-752 and the risk-informed methods the NRC approved to be acceptable to satisfy the requirements of 10 CFR Part 50, Appendix B. Additionally, I believe the NRC position is now clear (i.e., we did not grant an exemption from 10 CFR 50, Appendix B, and the reduction in commitment that we approved is acceptable to satisfy the requirements of 10 CFR 50, Appendix B), as evidenced by the inclusion of additional details related to the applicability of 10 CFR 50, Appendix B in similar requests from other licensees requesting to implementing Code Case N-752 at their facilities.

Concern D: Review of regulatory alignment and implications for NRC practices.

As expressed by the Panel, this concern pertains to the NRC approval to implement Code Case N-752 at ONS contradicts the Commissions policy because an approval under the 10 CFR 50.55(z) alternative process bypasses 10 CFR 50.69(b)(2), which requires a license amendment to recategorize SSCs and change special treatment for safety-related SSCs. In their report, the Panel described that the Commission expects the intent of its regulations to be met and normally this requires conforming to the regulations as stated.1 The Commission explained that if another regulation in Part 50 provides for specific exemption relief, or for 1 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50765 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

alternative methods of compliance, the criteria of the specific regulation are the appropriate considerations.2 The Commission stated that if the specific exemption criteria, or the alternative methods of compliance, can be satisfied, there is no need also to satisfy the criteria of 10 CFR 50.12(a).3 The Panel concluded that to assure that the staff is meeting the Commissions statement that when another regulation in Part 50 provides for specific exemption relief, or for alternative methods of compliance, the criteria of the specific regulation are the appropriate considerations,4 the staff should assure that the Commission is aware of, and approves of, the usage of the ASME Code Case N-752 50.55a(z) alternative. The Panel recommended that the staff should engage the Commission before continuing the practice of utilizing 10 CFR 50.55a(z)(1) to authorize alternatives that allow a licensee to use processes similar to processes set forth elsewhere in the Commissions regulations.

I disagree with the Panels conclusion. I do not agree that we needed Commission approval to approve the alternative requests and amendments to reduce commitments in the licensees QA programs for the LSS components addressed in Code Case N-752. While I agree that 10 CFR 50.69 is a separate and more robust regulation to recategorize SSCs for entire systems and change special treatment for safety-related SSCs, it is also voluntary, and I think Code Case N-752 provides appropriate flexibilities to allow licensees to complete timely repair/replacement of LSS components to support safe operations, while ensuring that the component remains capable of performing its safety-related function under design basis conditions. Code Case N-752 includes elements of defense in depth that are not in place for similar components treated under 10 CFR 50.69. The NRC staffs evaluation justified that the categorization process demonstrated that the defense-in-depth philosophy was maintained with regards to system redundancy, independence, and diversity. In addition, the NRC staff determined that the approvals were allowed under NRC regulations and aligned with Commissions direction in SRM-SECY-19-0036, application of the single failure criterion to NuScale actuation block valves, which directs the staff to 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 as part of any licensing review or other regulatory decision. Therefore, I am not directing any further action in response to Concern D.

Response to Recommendations Panel Recommendation 1: The NRC should examine the base set of standards identified through the audit process and meetings with the licensee, with industry input, as part of a generic review of N-752, to ensure that safety is maintained.

I partially agree with Recommendation 1. As discussed in my assessment of Concern B, while I do see the benefit in NRC providing generic approval of Code Case N-752, I disagree that additional constraints beyond those already included in the code case are necessary. I see no evidence that other licensees who choose to request approval to implement Code Case N-752 2 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

3 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

4 Specific Exemptions; Clarification of Standards, 50 Fed. Reg. 50764, 50775 (Dec. 12, 1985) (available at https://archives.federalregister.gov/issue_slice/1985/12/12/50761-50778.pdf#page=4).

would depart from the specified requirement to using the QA Programs and procedures that ONS clarified that they would follow. In addition, all NRC licensees have similar NRC-inspected corrective action programs and design change control processes that are required to be followed.

For this reason, I am directing DEX to develop a plan within 60 days for timely generic endorsement of Code Case N-752 that makes maximum use of the experience gained in prior approvals and is consistent with this Directors Decision, which is a culmination of significant staff resources expended evaluating both the safety and regulatory issues at play. The endorsement is an opportunity to highlight the important role that the QA change control process has in addressing potential interactions between passive and active features and common mode failures, as discussed in my response to Recommendation 2. As part of its generic endorsement, the NRC staff should ensure that all stakeholders are engaged to ensure common understanding of the issues.

Panel Recommendation 2: As part of a generic review of N-752, address the categorization/evaluation processes for LSS components to account for potential cascading failures and passive-active interactions in redundant systems.

I disagree with Recommendation 2. Generic endorsement of Code Case N-752 should be limited in scope as described in response to Recommendation 1. As I described in my assessment of Concern B, additional elements of defense in depth provided for in Code Case N-752 beyond what is required under 10 CFR 50.69 help to provide confidence there is low potential for inadvertent common cause failures to be introduced through interactions between passive and active components.

While I do see the benefit in NRC providing generic approval of the code case, I disagree that additional constraints beyond those already included in the code case are necessary. While plausible, I find that the coexistence of a risk significant simultaneous cascading common cause failure because of use of Code Case N-752 to be speculative and at best anecdotal and have confidence that licensees QA change control progress which we routinely inspect would evaluate any potential impact to active functions resulting from the use of alternative treatments of passive components. Directly applicable operating experience would be needed to support data-driven decision-making to justify considering additional regulatory expectations. I also see no evidence that other licensees who choose to request approval to implement Code Case N-752 would depart from the specified requirement to using the QA Programs and procedures that ONS clarified that they would follow. In fact, recent data indicates the opposite is true; subsequent licensee submittals have included additional clarity regarding applicability of 10 CFR 50, Appendix B in their alternative requests. All NRC licensees have similar NRC-inspected corrective action programs and design change control processes that are required to be followed. I do see our endorsement of the code case as an opportunity to highlight the important role that the QA change control process has in addressing potential interactions between passive and active features and common mode failures.

Panel Recommendation 3: The NRC may consider monitoring the industrys application of N-752s application to evaluate its practical impacts and ensure ongoing safety as its use expands. This could be accomplished through inspection, generic communications, or voluntary initiatives.

I agree with Recommendation 3. I agree that the NRC will continue to monitor industry implementation of this new program through our routine baseline inspection program. As I described in response to Concern A, I do see some benefit to providing some additional clarity to inspection staff to ensure we are taking an appropriate risk-informed and performance-based approach to oversight of licensees implementation of this new program. Based on the low safety significance of the components addressed by Code Case N-752, any inspection should be conducted using existing baseline inspection samples (i.e., resource neutral) and should not be prioritized over more significant inspection samples. Any inspection should focus on the actual performance of any SSCs receiving alternative treatments and associated ongoing required performance monitoring and corrective actions versus attempting to inspect the appropriateness of codes selected by licensees.

For this reason, I am directing DRO to provide recommended limited adjustments to relevant inspection procedures, as appropriate, within 6 months with a minimal level of NRC resources.

Concluding Remarks The DPO submitters positions were of notable technical merit and well documented in the submittal. I commend all of you for your commitment and dedication to the NRCs mission. I want to thank the DPO submitters for raising this DPO and for their active participation throughout the process. Their willingness to raise concerns is admirable and vital to ensuring a healthy safety culture within our agency. The discussions early in the process were incredibly valuable in ensuring that the issues and the safety implications were appropriately understood. I also want to thank the Panel for their thoughtful assessment of the concerns raised by the DPO submitters and for reaching a conclusion that ONS remains safe under the current use of their authorized alternative based on Code Case N-752 supported by the plants existing operational oversight mechanisms, including risk assessments, audits, and OpE programs.

The differing views on the use of Code Case N-752 has and will lead to significant improvements to clarity and reliability of its implementation. Specifically, the previously issued Information Notice 2025-01 provided clarity regarding the applicability of 10 CFR 50, Appendix B to passive LSS components. Improvements to the inspection and oversight program guidance will help ensure the agency has an appropriate risk-informed and performance-based regulatory footprint for reliable ongoing verification of compliance. Finally, a timely generic endorsement of Code Case N-752 will help to ensure implementation consistent with the NRC staffs evaluation and conclusions of the plant-specific approvals to date and the decisions in response to the differing views. My decision is informed by Commission direction to apply risk-informed principles for any licensing or regulatory decision and to regulate in a performance-based manner, ADVANCE Act direction to ensure impacts on agency decisions are commensurate with the safety significance of the differing opinion, and the NRC Principles of Good Regulation.

A summary of the DPO will be included in the Weekly Information Report (when the case is closed) to advise employees of the outcome.