RA-23-0282, 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

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

ML23293A267
Person / Time
Site:	Oconee
Issue date:	10/20/2023
From:	Ellis K Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation, NRC/RGN-II, Document Control Desk
References
RA-23-0282
Download: ML23293A267 (1)

Text

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

Emergency Preparedness Duke Energy 13225 Hagers Ferry Rd., MG011E Huntersville, NC 28078 843-951-1329 Kevin.Ellis@duke-energy.com RA-23-0282 10 CFR 50.55a October 20, 2023 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Duke Energy Carolinas, LLC Oconee Nuclear Station (ONS), Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55

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 By letter dated July 27, 2022 (Agencywide Document Access and Management System (ADAMS) Accession No. ML22208A031), as supplemented by letter dated March 9, 2023 (ADAMS Accession No. ML23068A015), Duke Energy Carolinas, LLC (Duke Energy) submitted a proposed alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, for Oconee Nuclear Station (ONS) Units 1, 2, and 3, and Keowee Hydro Station, Units 1 and 2. Specifically, Duke Energy requested to use the alternative requirements of ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1, for determining the risk-informed categorization and for implementing alternative treatment for repair/replacement activities on moderate and high energy Class 2 and 3 items in lieu of certain ASME Code Section XI, paragraph IWA-1000, IWA-4000, and IWA-6000 requirements.

From August 1 to October 10, 2023, the U.S. Nuclear Regulatory Commission (NRC) staff conducted a regulatory audit of the proposed alternative (ADAMS Accession No. ML23219A140). Subsequent to the regulatory audit, the NRC staff determined that additional information is needed to complete their review. Duke Energy received the request for additional information (RAI) from the NRC through electronic mail on October 11, 2023 (ADAMS Accession No. ML23284A332).

The enclosure provides Duke Energys response to the RAI questions.

No regulatory commitments are contained in this submittal.

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

r. DUKE

• ~ ENERGY

RA-23-0282 Page 2 Kevin M. Ellis General Manager - Nuclear Regulatory Affairs, Policy & Emergency Preparedness

Enclosure:

Response to Request for Additional Information cc:

Ms. Laura Dudes, Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. Shawn Williams, Senior Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, Maryland 20852 Mr. Jared Nadel NRC Senior Resident Inspector Oconee Nuclear Station

RA-23-0282 Enclosure Page 1 of 35 ENCLOSURE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

RA-23-0282 Enclosure Page 2 of 35

=

Background===

Note: U.S. Nuclear Regulatory Commission (NRC) text is italicized.

By letter dated July 27, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22208A031), as supplemented by letter dated March 9, 2023 (ML23068A015), Duke Energy Carolinas, LLC (Duke Energy, the licensee) submitted a licensing action request to authorize a proposed alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Specifically, the licensing action proposes to use the alternative requirements of ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1, (Code Case N-752) for determining the risk-informed categorization and implementing alternative treatment for repair/replacement activities on moderate-and high-energy Class 2 and 3 items instead of ASME Code Section XI, paragraphs IWA-1000, IWA-4000, and IWA-6000 requirements.

From August 1 to October 10, 2023, the U.S. Nuclear Regulatory Commission (NRC) staff conducted a virtual audit to support its review of the alternative. The audit plan was issued on July 3, 2023 (ML23178A068). The purpose of the audit was for NRC staff to gain a better understanding of the issues identified in the draft request for additional information (RAI) in the Enclosure to the Audit Plan and to identify information that may require docketing to support the basis of the NRC staffs licensing decision. Duke Energy established a Web-based portal (portal) to add documents needed to support audit discussions.

Based on the audit, the NRC staff has determined that additional information is needed.

RAI No. 1 Please provide the information presented during the audit that addressed the draft RAI No. 1 included as an Enclosure to the audit plan. The information related to this RAI was discussed during an audit call on August 1, 2023, and provided on the portal.

Duke Energy Response to RAI No. 1 Draft RAI No. 1 from Enclosure 2 of the Audit Plan is as follows:

In its relief request, Duke Energy proposes to implement the risk-informed categorization and treatment requirements of ASME Code Case N-752 when performing repair/replacement activities on Class 2 and 3 pressure-retaining items or their associated supports. Duke Energy stated that Code Case N-752 employs a comprehensive categorization process requiring input from both a PRA model and deterministic insights. It further stated that:

This approach will enable Dukes evaluation, categorization, and implementation of alternative treatments for resolution of emergent issues in segments of piping having low safety significance. Use of Code Case N-752 will also allow Duke Energy to identify and more clearly focus engineering, maintenance, and operations resources on critical components with high safety-significance, thus, enabling Duke Energy to make more informed decisions and increase the safety of the plant.

RA-23-0282 Enclosure Page 3 of 35 Duke Energy requests approval on the basis that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1). Duke Energy stated that The categorization and treatment requirements of Code Case N-752 are consistent with those in 10 CFR 50.69. The NRC staff evaluated Duke Energys plant-specific relief request to assess the sufficiency of the proposed use of Code Case N-752, for repair/replacement of Class 2 and 3 pressure-retaining items or their associated supports, employs a comprehensive categorization process. The NRC staff further evaluated whether the proposed use of Code Case N-752 would provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

The NRC staff notes Code Case N-752 is similar to ANO2-R&R-004 (ML090930246). The NRC issued safety evaluations approving adoption of 10 CFR 50.69 license amendments requests for implementing risk-informed categorizations to include the use of ANO2-R&R-004 (ML090930246) to categorize passive components within the 10 CFR 50.69 methodology framework. NRC Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, provides a risk-informed method for establishing the requirements for treatment of SSCs (systems, structures and components) by categorizing SSCs according to their safety significance. RG 1.201 endorses NEI 00-04, 50.69 Categorization Guidelines. The staff notes that there are some variances in Code Case N-752 with regard to incorporating the expected critical requirements of the methodology, as approved in the integrated decision-making process outlind in RG 1.201 and 10 CFR 50.69.

The NRC has concern that Code Case N-752s approach to risk-informed SSC categorization does not address the following attributes:

1. Code Case N-752 allows categorization on a component basis. As such, it appears that only deterministic insights are used to verify the impact of the passive component to other components in the system and function(s) the component(s) supports (active vs.

passive functions). This may, for example, categorize passive features of components as low safety significant (LSS) if the active features of the component within the system are determined to be high safety significant (HSS). Based on this variance, it is not apparent that the plant-specific categorization and treatment using Code Case N-752s would maintainan acceptable level of quality and safety at the component and system level.

Duke Energy Response to Draft RAI No. 1, Item 1 from Enclosure 2 of the Audit Plan:

While it is true that ASME Code Case N-752 allows categorization to be performed for individual components, additional insights other than deterministic are used to verify the impact of the passive component to other components. The impact of passive component failures on other components has been demonstrated by the use of the Consequence Evaluation methodology contained in ASME Code Case N-752 at numerous plant sites in a variety of NRC approved applications (e.g. Risk Informed Repair/Replacement Activities (RI-RRA), Risk-Informed Inservice Inspection (RI-ISI), 10CFR50.69). Please see References 1 through 8, as well as the response to Item 2 below.

As required by ASME Code Case N-752 and as discussed in the NRC Safety Evaluation for approval of ASME Code Case N-752 at Arkansas Nuclear One (Reference 1), in EPRI Topical Report TR-112657 Rev B-A (Reference 5), and in the Topical Reports pilot plants, the Consequence Evaluation must address not only the postulated failure of the subject pressure

RA-23-0282 Enclosure Page 4 of 35 boundary component (e.g. loss a flow path) but also other direct and indirect effects associated with the pressure boundary failure. For example, ASME Code Case N-752 requires an evaluation of the impact of the postulated failure on other active and passive components and functions that may be lost or impaired by the postulated pressure boundary component failure.

These impacts may include but are not limited to valves failing to open to allow required flow, valves failing to close thereby diverting flow from its intended target, pumps failing to provide flow to required loads, pumps failing to trip thereby failing to preserve required inventory, and failure of electrical equipment (e.g. due to spray or flooding) to provide motive power or control logic. These steps of the process are described in ASME Code Case N-752, Section I-3.3.1, Failure Modes and Effects Analysis. This analysis (I-3.3.1(a) through (g)) is inherent in the PRA for components important to risk.

After all the impacted equipment (failed or impaired, direct and/or indirect) is identified, the PRA model is revised and re-quantified, as necessary, crediting only the unaffected equipment to obtain Conditional Core Damage Probability (CCDP) and Conditional Large Early Release Probability (CLERP) values. Specifically, the quantitative steps are described in ASME Code Case N-752 Steps:

I-3.3.2 (a) (3) for the Initiating Event Impact Group Assessment I-3.3.2 (b) for the System Impact Group Assessment I-3.3.2 (c) for Combination Impact Group Assessment I-3.3.2 (d) (1) for Containment Performance Group Impact Assessment To further clarify:

ASME Code Case N-752 Appendix I provides guidance on assigning CCDP/CLERP consequence categories to segment failures on the basis of the number of available (i.e.,

unaffected by the rupture) mitigating trains remaining, broad categories of initiating event frequencies, and exposure times.

These remaining mitigating trains may be parallel system trains or other systems that provide a backup function to the unavailable system and that are unaffected by the direct and indirect consequences of the segment rupture.

The methodology to assign segment failures to consequence categories is based on the number of unaffected trains available to mitigate an event.

The specific decision criteria used to determine the consequence category depends on the type of impact the segment failure has on the plant and the reliability of the unaffected trains.

RA-23-0282 Enclosure Page 5 of 35 Figure 1 outlines how the unavailability interval (reliability) correlates to the backup train worth in ASME Code Case N-752 Table I-2.

Given a segment failure and all the associated spatial effects, the CCDP is the probability that the resulting scenario will lead to core damage (similarly for CLERP).

If the failure of a segment is estimated to lead to a core damage event with a probability greater than 1E-4, the segment is categorized as High consequence. An estimated CCDP within the range of 1E-6 to 1E-4 is categorized as Medium consequence. CCDPs less than 1E-6 are categorized as Low consequence.

This same Consequence Evaluation methodology has been used, coupled with probability of failure, since 1998 to evaluate the safety significance of pressure boundary failures for the Risk Informed In-Service Inspection program (RI-ISI). The RI-ISI methodology was first approved by NRC for Vermont Yankee in a RI-ISI application applied to Class 1 piping on November 8, 1998 (Reference 2). It was subsequently approved on December 29, 1998 for ANO Unit 2 in a RI-ISI application applied to Class 1, 2, 3 and non-nuclear safety piping (Reference 3). These pilot plant applications led to the NRC generic approval of Topical Report EPRI TR-112657 Rev B-A on October 28, 1999 (Reference 5) which has been codified as ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Appendix R, Supplement 2 which has been endorsed in 10CFR50.55a (Reference 8). Section 3 of EPRI TR-112657 Rev B-A explains how the consequence of failure methodology is applied and the basis for its use including the CCDP/CLERP thresholds. The accompanying NRC Safety Evaluation which discusses NRCs acceptance of the Consequence Evaluation methodology was dated October 28, 1999.

In the ASME Code Case N-752 categorization methodology, the failure of the components pressure boundary function is assumed with a probability of 1.0 and only the consequence evaluation is performed. The consequence evaluation is coupled with additional deterministic considerations (e.g., DID, safety margins) required by ASME Code Case N-752 Section I-3.4.2 in determining safety significance. The same steps are applied in the 10 CFR 50.69 passive component categorization methodology.

Alternate Treatment requirements (including repair/replacement) affect only the frequency of passive component failure. Categorizing solely based on consequences, which measures the safety significance of the pipe given that it ruptures, is conservative compared to including the rupture frequency in the categorization. The categorization will not be affected by changes in frequency arising from changes to the treatment. As recently as December 15, 2022 (Reference 15), the NRC agreed the consequence evaluation appropriately categorizes passive components, the method is conservative since it uses a 1.0 probability of failure, and that the changes in treatment only affect probability of failure.

As such, and as exemplified by numerous NRC precedents, the plant-specific categorization and treatment using ASME Code Case N-752 maintains an acceptable level of quality and safety at the component and system level.

Active HSS and Passive LSS Some components (e.g. valves) have both active and passive (pressure retaining) functions. It is possible when applying the 10 CFR 50.69 process for an SSC to have an active HSS and passive LSS categorization. This question was addressed in the Vogtle pilot plant review for 10 CFR 50.69 implementation in RAI 29 (Reference 9) and specifically discussed in the NRC

RA-23-0282 Enclosure Page 6 of 35 Safety Evaluation for that application (Reference 6). The Vogtle RAI and Vogtle response are provided below with minor edits for clarification underlined.

Vogtle RAI #29 (Reference 9):

RAI: Clarifications 2 and 3 in Section 3.1.3 of the LAR indicate that a passive component (i.e., a pressure retaining component) whose failure could fail an active HSS function can be assigned to LSS if the passive categorization process yields an LSS category.

The passive categorization is driven by the consequence of failure and not the frequency so it is unclear how the pressure retaining function of a passive component whose failure would fail an HSS function can be found LSS in the passive categorization process.

Please explain and provide an example.

Vogtle Response (adapted herein for Oconee): 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. Active risk is determined using insights from the PRA and other qualitative considerations. Passive risk is determined using a passive component categorization methodology. Risk associated with defense in depth is determined using guidance provided in the NEI 00-04 categorization methodology. The final risk of a component is the highest of these three risks. Then the critical attributes are identified for each HSS components to further understand the reason(s) for being HSS. For example, an HSS Motor Operated Valve (MOV) may have a critical attribute of fail to close because that is what made it HSS. However, the same valve may be LSS for passive risk (i.e., pressure boundary retention) assuming there is sufficient redundancy to respond to the event of interest and LSS from a defense in depth evaluation.

Therefore, It is possible to have an LSS passive component that fails an HSS function if there is sufficient redundancy to respond to the event of interest. That is, there are other unaffected components/trains/systems available to fulfill the function.

It is not possible to have an LSS passive component that fails an HSS function if there is not sufficient redundancy to respond to the event of interest.

As the Vogtle RAI states, the passive categorization process is driven by the consequence of failure in that the process conservatively assumes that a failure occurs with a probability of 1.0. As such, some postulated passive failures will be categorized as HSS while from a pure risk perspective they may be low safety significant. As an example, postulated failures with conditional core damage probability (CCDP) values of 5 E-04 are HSS per the passive categorization process. However, many passive components have failure frequencies of 1E-08 and lower. Thus, if failure frequency were to be considered, they may be shown quantitatively to be low safety significant.

As an illustration, the following is provided as a generic example and does not represent Oconee specific systems. As shown in the figure below, for the RCS heat removal function, MFW, EFW, AFW or HPSI can individually fulfill that function. Note also that EFW and HPSI for example, are two train systems. From an active risk perspective, we will assume that the PRA risk insights or other qualitative considerations have determined the HPSI pumps to be HSS. However, if we were to postulate a pressure boundary failure that failed one HPSI pump, and no other impacts (e.g. spatial, loss of inventory), we see from the figure that not only is the second HPSI pump train available

RA-23-0282 Enclosure Page 7 of 35 to fulfill the RCS heat removal function, but MFW, EFW and AFW are available to fulfill the RCS heat removal function; therefore, this passive failure of one HPSI pump would be considered LSS.

Now from a different perspective, consider the following example in which a two train system is supplied from a tank. One of the system functions is to draw water from the tank and deliver downstream. Assume that the PRA risk insights or other considerations identify Pumps A and B as HSS. Therefore, all active components that support the function are preliminary HSS.

Reaetivityt RCS Heat Removal MFW EFW ~-------+t RPS AFW LTOP mitiaia HPSI.

EOCS Agure3-3 RCS Long Tenn lnvento'Y lnvento,y : Control/Heat Removal Control

. P~N,--------.

HPSI HPR css

.,......_....... I css IUI ~~

Heat Removal, lnwn10fY Control, and Long tenn HNl Removal Safety FullCtlon*

RA-23-0282 Enclosure Page 8 of 35 In order to determine risk of the passive components (i.e., pressure retention components), passive component categorization methodology is used. This methodology divides the entire system into(segments do not, necessarily, have to be small - but must have the same consequence) segments and postulates failure of one segment at a time. The following discussion outlines, at high level, the potential outcome from the passive component categorization perspective and compares with a component/function risk which was determined to be preliminary HSS from an active risk perspective.

Point X only:

A failure of a passive component (i.e., pipe) at point X would prevent Train A and Train B from performing its function. In contrast to the previous example, we assume there are no other unaffected trains/systems to fulfill this function, therefore, the passive component at Point X (i.e., pipe) would be candidate HSS (see ASME Code Case N-752 Table I-2). This would complement with the component/function risk from an active risk perspective. Note that the isolation valves pressure boundary failure is part of this segment failure as it is also unisolable and prevents isolation of at Points Y and Z.

Point Y only:

A failure of a passive component (i.e., pump) at point Y would prevent Train A from performing its function at the train level; however, Train B is not affected, assuming the isolation valve can be closed in time (requires detection, time, procedures, etc.) so Train B would be able to perform the function at the train level, which in turn, will fulfill the function at system level. Hence, the passive component at Point Y (i.e., pump) could be candidate LSS for pressure retention function although it is in the path of an HSS Tank Pointy Point.X Train A Tr-ainB

RA-23-0282 Enclosure Page 9 of 35 function. The ultimate passive categorization rank would be a function of frequency of challenge and exposure time to challenge (see ASME Code Case N-752 Table I-2). It should be noted that the final risk will be the higher of active risk, passive risk, or defense in depth. So if the pump was determined to be HSS in any of these risk evaluations, then its final categorization would be HSS.

However, as discussed in industry 10 CFR 50.69 LARs and allowed by Regulatory Guide 1.201, following the guidance in section 10.2 of NEI 00-04 (Detailed SSC Categorization) would show the pressure boundary function of the pump to be LSS and allow alternative treatments to be applied to the pressure retaining portions of the component.

Point Z only:

A failure of passive component (i.e., pipe) at point Z would prevent Train B from performing its function; however, Train A is not affected, assuming the isolation valve can be isolated in time (requires detection time, etc.) so Train A would be able to perform the system function. Hence, the passive component at Point Z (i.e., pipe) would be candidate LSS although it is in the path of an HSS function. Similar to point Y, the ultimate passive categorization rank would be a function of frequency of challenge and exposure time to challenge. (see ASME Code Case N-752 Table I-2)

As demonstrated above, it is possible to have an LSS passive component that fails an HSS component provided that there are a sufficient number of unaffected backup trains available to fulfill the function. However, postulated passive component failures that fail a function (e.g. zero redundancy) will be categorized as HSS.

The NRC staff accepted this explanation in the Vogtle 10 CFR 50.69 Safety Evaluation (Reference 6).

The discussion above refers to 10 CFR 50.69 activities, however the ASME Code Case N-752 categorization process is identical to the 10 CFR 50.69 methodology for pressure boundary components, captures identical system impacts, and results in the same conclusion for passive/pressure boundary components. The ASME Code Case N-752 categorization process considers the direct and indirect effects of the pressure boundary component failure regardless of the specific categorization scope (i.e. the in scope break may affect out of scope components) and, therefore, produces the same results whether a single component is categorized or an entire system is categorized.

The ASME Code Case N-752 process provides a well-structured approach for determining the risk significance of pressure boundary failures including defense in depth considerations and impacts of the break on active components/functions thereby providing an acceptable level of quality and safety.

Draft RAI No. 1, Item 2 from Enclosure 2 of the Audit Plan:

2. Section I-3.4.2(b)(1) of Code Case N-752 requires analysis and assessment of whether the failure of the pressure retaining function of the segment will directly or indirectly (e.g.,

through spatial effects) fail a basic safety function. Given that the active components within the system also play a significant role to establish the basic safety functions, an assessment of the risk significance of these active components would provide reasonable assurance that an acceptable level of quality and safety be maintained. The

RA-23-0282 Enclosure Page 10 of 35 NRC staff noted that Code Case N-752 requires only Conditional Core Damage Probability (CCDP) (which can be estimated based on RAW values) as the risk screening criteria and conditional large early release probability (CLERP), given a specific failure (e.g., piping segment failure). Both RAW and Fussell-Vesely (F-V) are risk importance measures that uniquely highlight component significance to address basic safety functions. RAW importance reflects the increase in a selected figure of merit when an SSC is assumed to be unable to perform its function due to testing, maintenance, or failure. It is the ratio or interval of the figure of merit, evaluated with the SSCs basic event probability set to one. F-V importance is the fractional contribution to the total of a selected figure of merit for all accident sequences containing that basic event. The F-V importance measure is calculated by determining the fractional reduction in the total figure of merit brought about by setting the probability of the basic event to zero. The objective in applying the F-V criterion is to ensure that any entity that has an unusually large contribution to risk is identified, regardless of CDF or LERF. Based on the above, it is not apparent that the Code Case N-752s plant-specific approach of using only CCDP and CLERP is sufficient as risk screening criteria to provide an acceptable level of quality and safety.

Duke Energy Response to Draft RAI No. 1, Item 2 from Enclosure 2 of the Audit Plan:

Traditionally, relative risk metrics such as RAW and FV have been used to risk rank active components (e.g. Maintenance Rule). Relative risk metrics are useful in identifying active components role in basic safety functions and the remaining layers of defense in depth to inform prioritizing activities (special or alternative treatments) applied to active functions (e.g.

valves that need to change state to fulfill a function). However, as discussed below, this is not the case for risk ranking/prioritizing pressure boundary components. Further, ASME Code Case N-752 does not change treatments of items (e.g. valve internals or valve operators) that perform active functions. ASME Code case N-752 is focused on the pressure boundary function and only treatments to the pressure boundary aspects of a component can be changed. Note:

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. Given that altering the active function treatment is NOT a goal of N-752, only potential changes to the passive pressure boundary, utilizing a consequence-based approach (e.g. setting the failure probability to 1.0) to risk assess the passive function does, in fact, provide an acceptable level of quality and safety for the purpose of risk assessing the pressure retaining functions.

As noted in Duke Energys response to Item 1 above, the ASME Code Case N-752 categorization methodology is the consequence evaluation portion of the EPRI TR-112657. As some background, EPRI TR-112657 Revision B-A is the foundational methodology for several risk-informed applications related to SSCs that perform pressure boundary functions. The methodology contained in EPRI TR-112657 Revision B-A is summarized in Figure 2 and consists of two main analysis steps. That is, a Consequence of Failure Evaluation step and a Failure Potential Evaluation step. A timeline for development, approval and use of TR-112657 and its daughter methodologies, e.g. ASME Code Case N-660 and N-752, is provided as Figure

3. Figure 3 also provides a timeline of other related industry efforts that present the background, development, and the technical robustness of the risk-informed categorization process contained in ASME Code Case N-752. Note: the timeline in Figure 3 also shows additional applications where the same consequence assessment methodology has been applied and approved by the NRC for pressure boundary risk assessments. Specifically, ASME Code Case N-660, which is referenced in RG 1.201, is noted as an acceptable method for

RA-23-0282 Enclosure Page 11 of 35 addressing the pressure-retaining function or passive function of active components for 10 CFR 50.69. ASME Code Case N-660 was accepted by the NRC staff in a 2005 revision of RG 1.147.

During the development of EPRI TR-112657 and related efforts, a review of relative risk metrics (e.g. NUREG/CR-3385) and other ways of risk ranking components was conducted. These efforts have shown the challenges in using relative risk metrics (e.g RAW, FV, RRW) for risk ranking of pressure boundary components. The challenges include but are not limited to:

Use of RAW is generally not an appropriate metric for component failures that result in initiating events are typically input as a frequency rather than a probability.

Use of RAW is not appropriate for components with very low failure rates whose failures result in initiating events. For example, as discussed in NRC approved Topical Report WCAP-14572 (Reference 10) dated February 1999 Piping failure probabilities are typically very small compared to other component failures modeled in the PSA. When the failure probability is set to 1.0 for the RAW calculation, large RAW values typically result. Therefore, the EPRI guideline classifying a segment as high safety-significant for RAW values greater than 2 does not provide meaningful results.

There exist large variations in the distribution of failures rates between individual pressure boundary components and postulated break sizes as well as system to system variations (see Tables 1 and 2). Use of relative risk metrics will show components/systems with very low failure rates as low safety significance as compared to other components/systems even if the postulated failure results in high consequences (e.g. limited or no defense in depth).

F-V, which is numerically equivalent to RRW, is generally not appropriate for components with very low failure rates. Challenges in using RRW is further discussed in the 2004 report (Reference 11) produced by the European Nuclear Regulators Working Group on RI-ISI. A summary of these concerns is provided as Figure 3.

Pressure boundary components typically have very low failure rates/probabilities for those breaks of most concern (e.g. large and very large breaks). As such, using relative importance measure such as RRW and F-V identifies the vast majority of pressure boundary components as low safety significant. See Figure 4, Figure 5, Figure 6.

The differences in failure rates/probabilities of concern for pressure boundary components as compared to active components is quite large (e.g. 5 to 12 orders of magnitude). Figures 4, 5 and 6 demonstrate that these large differences in failure rates/probabilities lead to very low safety significance assignment for the vast majority of pressure boundary components, including a large fraction of the reactor coolant pressure boundary when utilizing relative importance measures for determining safety significance.

Use of the CCDP and CLERP metrics avoids these identified concerns from previous efforts.

For example, uncertainties associated with pressure boundary components that have very low failure probabilities are eliminated since failure is assumed (i.e. a failure probability of 1.0 is used). Additionally, use of the CCDP and CLERP metrics, which includes identifying all active functions that are and are not impacted by the postulated pressure boundary failure, identifies pressure boundary components as important from a defense in depth perspective if there is limited or no redundancy given the postulated pressure boundary failure even if the postulated

RA-23-0282 Enclosure Page 12 of 35 pressure boundary failures overall contribution to risk (CDF/LERF) is extremely low. See Figure 4 and Figure 5. Therefore, functions of active components which are failed by the postulated break are addressed in the process.

As documented in numerous NRC Safety Evaluations, the ASME Code Case N-752 process provides a well-structured approach for determining active and passive components impacted by the postulated break and the resulting risk significance of these impacts thereby providing an acceptable level of quality and safety.

Draft RAI No. 1, Item 3 from Enclosure 2 of the Audit Plan:

3. Code Case N-752 Table I-2 provides a defense-in-depth matrix for evaluating consequence categories and frequency of events using a quantitative matrix of CCDP.

Code Case N-752 however, allows either risk screening or defense-in-depth as noted in Section I-3.3.2 (a) (b) and (c): Differences in the consequence rank between the use of Table I-1 [Table I-2, Table I-3] and Table I-5 shall be reviewed, justified and documented or the higher consequence rank assigned. Both defense-in-depth and risk screening criteria are essential for evaluating SSC categorization. Based on the above, it is not apparent that the Code Case N-752s plant-specific approach would provide an acceptable level of quality and safety.

Duke Energy Response to Draft RAI No. 1, Item 3 from Enclosure 2 of the Audit Plan:

Duke Energy agrees with the NRC staff in that both defense in depth and the risk evaluation are essential in developing a risk-informed categorization for systems and individual components.

As noted in response to Item 1 above, the risk screening criteria inherently addresses defense-in-depth and results from the quantitative evaluation are compared to the results from the qualitative equivalent assessment using the applicable tables in ASME Code Case N-752.

Note: The process does not allow only consideration of the quantitative CCDP/CLERP result but requires comparison to the qualitative evaluation. Specifically, ASME Code Case N-752 Section I-3.3.2(a)(3) states:

Differences in the consequence rank between the use of Tables I-1 and Table I-5 shall be reviewed, justified and documented or the higher consequence rank assigned.

Similar statements are in the System, Combination, and Containment Performance Impact Group Assessment sections. Further, I-3.4.2(b) contains a list of additional Categorization Considerations that must be met or HSS is assigned.

As some background, section I-3.3.2 of ASME Code Case N-752 was developed from section I-3.1.2 of ASME Code Case N-660 which was approved by ASME in 2002 and accepted in Regulatory Guide 1.147 in 2005. These tables are the same in both ASME Code Case N-660 and ASME Code Case N-752. At that time, the ASME/ANS PRA Standard had not been developed and Regulatory Guide 1.200 did not exist. As such, there was some concern with the level of robustness of some plant-specific PRAs. Tables I-1, I-2 and I-3 of ASME Code Case N-660 were developed as a means to calibrate plant-specific results as well as assuring adequate levels of defense in depth. Examples of how each of the tables are used is provided as follows with the understanding that ASME Code Case N-660 is from 2002:

Example 1: The segment has a postulated pipe break that results in small LOCA but does not disable any backup trains and has a PRA calculated CCDP of 1E-07 (i.e. low

RA-23-0282 Enclosure Page 13 of 35 consequence rank per Table I-5). The user would have to compare that result to Table I-1 which recommends that small LOCAs be assigned either a high or medium consequence rank. The user would then be required to develop the technical basis (i.e.

additional defense in depth) on why a low consequence rank is justifiable or assign the postulated failure to a high consequence rank unless a medium consequence rank could be justified.

Example 2: The segment has a postulated pipe break in a system designed to respond to a reactor trip (anticipated event), with all year exposure time (standby system that is tested only once per year),),), that also fails all other systems used to respond to a reactor trip, except one backup train, results in a PRA calculated CCDP of 1E-07 (i.e.

low consequence rank per Table I-5). The postulated break does not cause an initiating event and only causes system/train loss. The user would have to compare that result to Table I-2 which recommends that for postulated breaks designed to respond to a reactor trip with only one backup train for those conditions and all year exposure time be assigned a high consequence rank. The user would then be required to develop the technical basis (i.e. additional defense in depth) on why a low consequence rank is justifiable or assign the postulated failure to a high consequence. Note: I-3.3.2(b) also states Additionally, for defense-in-depth purposes, all postulated failures leading to zero defense (i.e., no backup trains) shall be assigned a high consequence. This is required regardless of any PRA calculations using Table I-5.

Example 3: The segment has a postulated pipe break that results in small LOCA as well as disabling all but one backup train results in a CCDP of 1E-07 (i.e. low consequence rank per Table I-5). The postulated break causes an initiating event and causes loss of a system/train. The user would have to compare that result to Table I-3 which recommends that small LOCAs with only one backup train available be assigned a high consequence rank. The user would then be required to develop the technical basis (i.e.

additional defense in depth) on why a low consequence rank is justifiable or assign the postulated failure to a high consequence rank.

With the significant increases in PRA Technical Adequacy over the years, in particular in the US, consideration was given to revising ASME Code Case N-752 to eliminate Table I-1, I-2 and I-3 and allow plants with robust peer reviewed PRAs to use Table I-5 directly. However, because the ASME Code is an international code and not all countries follow the ASME/ANS PRA standard, those changes were not made. Further, leaving those requirements in place for US users of ASME Code Case N-752 provides an additional layer of confidence that the ASME Code Case N-752 process will produce technically robust results and is consistent with risk-informed decision-making philosophy. The notion that the ASME Code Case N-752 process provides technical robust results is supported by the code cases endorsement by several standards development organizations, technical support organizations, as well as several international regulatory bodies.

As noted in the original relief request submittal, Oconee PRA models have been peer reviewed and there are no PRA upgrades that have not been peer reviewed. The PRA models credited in the ASME Code Case N-752 application are the same PRA models credited in the License Amendment Oconee Nuclear Station, Units 1, 2, And 3, Issuance of Amendments Regarding Adoption of Technical Specification Task Force (TSTF) -425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -Risk-Informed Technical Specification Task Force (RITSTF) Initiative 5b," and License Amendment "Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based

RA-23-0282 Enclosure Page 14 of 35 Fire Protection Program in Accordance with 10 CFR 50.48(c)," with routine maintenance updates applied. As such, the Oconee PRA models are an example of the significant improvements in PRA Technical Adequacy that supports use of ASME Code Case N-752 Table I-5 directly. However, as noted throughout this document use of ASME Code Case N-752 Tables I-1, I-2, and I-3 will continue to be used as a comparison to the results from Table I-5.

Thus, the ASME Code Case N-752 process provides an acceptable level of quality and safety due to the aforementioned discussion regarding the use of defense in depth and a quantitative risk evaluation in categorizing components.

References:

1. U.S. Nuclear Regulatory Commission, Arkansas Nuclear One, Units 1 and 2 - Approval of Request for Alternative from Certain Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (EPID L-2020-LLR-0076),

dated May 19, 2021 (ADAMS Accession No. ML21118B039).

2. Letter from U.S. Nuclear Regulatory Commission to Vermont Yankee Nuclear Power Corporation, Request to Use Code Case N560 as an Alternative to the Requirements of ASME Code,Section XI, Table IWB-2500-1 at Vermont Yankee Nuclear Power Station (TAC No. M99389), dated November 9, 1998 (ADAMS Accession No. ML20195C403; enclosed Safety Evaluation at ADAMS Accession No. ML20195C416).

3. U.S. Nuclear Regulatory Commission Letter to Entergy Operations, Inc., Request to Use a Risk-Informed Alternative to the Requirements of ASME Code Section XI, Table IWX-2500 at Arkansas Nuclear One, Unit No. 2 (TAC No. M99756), dated December 29, 1998 (ADAMS Accession No. ML20198M762; enclosed Safety Evaluation at ADAMS Accession No. ML20198M784).

4. Letter from Dana A. Powers (Chairman, Advisory Committee on Reactor Safeguards) to Dr. William D. Travers (Executive Director for Operations, U.S. Nuclear Regulatory Commission), Safety Evaluation Report Related to Electric Power Research Institute Risk-Informed Methods to Inservice Inspection of Piping (EPRI TR-112657, Revision B, July 1999), dated September 15, 1999 (ADAMS Accession No. ML992650028).

5. Electric Power Research Institute (EPRI) TR-112657, Rev B-A Revised Risk-Informed Inservice Inspection Evaluation Procedure (PWRMRP-05), dated December 1999 (ADAMS Accession No. ML013470102).

a. Including the NRC Safety Evaluation dated October 28, 1999. ML993190474

6. Letter from U.S. Nuclear Regulatory Commission to Southern Nuclear Operating Company, Inc., Vogtle Generating Electric Plant, Units 1 and 2 - Issuance of Amendments RE: Use of 10 CFR 50.69 (TAC Nos. ME9472 and ME9473), dated December 17, 2014 (ADAMS Accession No. ML14237A034).

7. U.S. Nuclear Regulatory Commission Letter to Tennessee Valley Authority, Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendment Nos. 317, 340, and 300 Regarding Adoption of Title 10 of the Code of Federal Regulations Section 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power plants, dated July 27, 2021 (ADAMS Accession No. ML21173A177).

RA-23-0282 Enclosure Page 15 of 35

8. Federal Register, American Society of Mechanical Engineers 2019-2020 Code Editions, 87 FR 65128 (October 27, 2022) (Codified at 10 CFR 50.55a).

9. Southern Nuclear Operating Company Inc. Letter to U.S. Nuclear Regulatory Commission, 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 (ADAMS Accession No. ML14122A364).

10. Westinghouse Non-Proprietary Class 3 Report WCAP-14572, Revision 1-NP-A, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report, dated February 1999 (ADAMS Accession No. ML042610469).

11. European Commission Report prepared by The Nuclear Regulators Working Group Task Force on Risk-Informed Inservice Inspection, Report on the Regulatory Experience of Risk-Informed Inservice Inspection of Nuclear Power Plant Components, dated August 2004.

12. Report by Nuclear Energy Agency Committee on the Safety of Nuclear Installations (CSNI) Integrity and Ageing Working Group (IAGE), EC-JRC/OECD-NEA Benchmark Study on Risk Informed In Service Inspection Methodologies (RISMET), dated November 2010.

13. Tennessee Valley Authority Letter to the U.S. Nuclear Regulatory Commission, Browns Ferry Nuclear Plant (BFN) - Unit 3 - Request for Approval of the BFN American Society of Mechanical Engineers (ASME)Section XI Alternate Inservice Inspection Program -

Risk Informed Inservice Inspection (RI-ISI) and Cost Beneficial Licensing Action (CBLA) 99-01, dated April 23, 1999 (ADAMS Accession No. ML18039A754; enclosed ISI Program is at ADAMS Accession No. ML18039A756).

14. FirstEnergy Nuclear Operating Company Letter to the U.S. Nuclear Regulatory Commission, Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Risk-Informed Inservice Inspection Program Plans ISI (Inservice Inspection) Program Relief Request, dated July 24, 2002 (ADAMS Accession No. ML022060549).

15. U.S. Nuclear Regulatory Commission Letter to Energy Northwest, Columbia Generating Station - Issuance of Amendment No. 269 RE: License Amendment to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors (EPID L-2021-LLA-0207), dated December 15, 2022 (ADAMS Accession No. ML22308A096).

RA-23-0282 Enclosure Page 16 of 35 Figure 1. Crediting Unaffected Backup Trains (Reference 5)

Number of Backup system / trains available To standardize crediting backup trains, a train "worth" concept is used (e.g.

A train of turbine driven EFW (unavailability of approximately 1 E-1) should not be credited equally as a train of motor driven EFW (unavailability of approximately 1 E 2).

- a backup train "worth" is defined to have a mean unavailability value of approximately 1 E-2.

611ckup*Tr11in*

"'Wonh"'o 0.5.o 1o 1.So 2*

2.S-3" System/Train FW HPCI RCIC 1 LPCI Train 2 LPCI Trains 1 CS Train 2 CS Trains U nav11il11 bility *lntervatl Unavailability-Mean*Valuea 3E-2*to*3E*1*

1E-1o 3E-J*to*3E-DI 1E-2o 3E-Ho*3E-J.o 1E.Jo 3E-5*to*3E-4*

1E-4o 3E,6*1o*3E,5" 1E-So 3E-7-lo*3E*6o 1E-6o Unavailability Corresponding Train Worth" 1.SE-2 1

8.8E-2 o.s 1.1E*1 0.5 9.2E*3 1

2.SE-4 2

1.1E*2 1

3.9E-4 1.5

RA-23-0282 Enclosure Page 17 of 35 Figure 2. RI-ISI Methodology Overview (Reference 5)

RI-ISi Methodolgy I Determine Scope I l

j Perform Consequence Perform Failure Poterntial Evaluation Evaluation l

l I

I Perform Service Review I Determine Segment Risk Category :

Adjust Select Elements for Inspection &

Element Performance Selection Element Inspection Methods Monitoring I

Perform Risk Impact Assessment I

I Finalize Program !

.u I

RA-23-0282 Enclosure Page 18 of 35 Figure 3. Timeline 1991 - American Society of Mechanical Engineers, "Risk-Based Inspection - Development of Guidelines, Volume 1, General Document," CRTD-Vol. 20-1, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C.

1996 - American Society of Mechanical Engineers, Case N-560, Alternative Examination Requirements for Class 1, B-J Piping Welds,Section XI, Division 1, August 9, 1996 1997 - American Society of Mechanical Engineers, Case N-578, Risk Informed Requirements for Class 1, 2, and 3 Piping, Method B,Section XI, Division 1, September 2, 1997 1997 - Letter, dated September 30, 1997, D. C. Mims (Entergy Operations, Inc.) to Document Control Desk (NRC), containing results of pilot plant study for risk-informed ISI program, letter submitting eight system evaluations, dated Sept 30, 1997. (ADAMS Accession No. ML20217E899) 1998 - Letter, dated March 31, 1998, D. C. Mims (Entergy Operations, Inc.) to Document Control Desk (NRC), requesting approval of a risk-informed alternative for examination of piping systems, letter submitting service water system evaluation dated March 31, 1998. (Cover Letter is at ADAMS Accession No. ML20217N833) 1998 - VY Safety Evaluation, Request to Use Code Case N560 as an Alternative to the Requirements of ASME Code Section XI Table IWB-2500-1 at Vermont Yankee Nuclear Power Station, - Issuance of Amendment (TAC No. 269 RE: M999389), dated November 9, 1998 (Cover Letter is at ADAMS Accession No. ML20195C403; the enclosed Safety Evaluation is ML20195C416)

Including RAIs and responses 1998 - ANO-2 Safety Evaluation, Request to Use a Risk-Informed Alternative to the Requirements of ASME Code Section XI, Table IWX-2500 at Arkansas Nuclear One, Unit No. 2 (TAC NO. M99756) ANO-2 SE dated December 29, 1998 (Cover Letter at ML20198M762 and enclosed SE at ML20198M784) including RAIs and responses 1998 - American Society of Mechanical Engineers, "Risk-Based Inspection - Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components," CRTD-Vol. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C. (searched - cannot locate in ADAMS) 1999 - ACRS Letter dated September 15, 1999, Dana Powers (Chairman, ACRS), to Dr, William Travers (USNRC), Safety Evaluation Report Related to Electric Power Research Institute Risk-Informed Methods to Inservice Inspection of Piping. (ADAMS Accession No. ML992650028)

ACRS meetings May, 1999 and September, 1999 1999 - SE on TR-112657 dated October 28, 1999 (Cover letter at ML993190460 and SE at ML993190474)

RA-23-0282 Enclosure Page 19 of 35 1999 - Electric Power Research Institute (EPRI) TR-112657, Rev B-A Revised Risk-Informed Inservice Inspection Evaluation Procedure, dated December 1999. (The A version of the Topical Report was transmitted to the NRC by letter dated 2/10/2000; ADAMS Accession No. ML013470102)

Including RAIs and responses 2002 - American Society of Mechanical Engineers, Case N-660, Risk Informed Safety Classification for use in Risk-informed Repair / Replacement Activities,Section XI, Division 1, July 23, 2002 2004 - Report on the Regulatory Experience of Risk-Informed Inservice Inspection of Nuclear Power Plant Components and Common Views, Prepared by The Nuclear Regulators Working Group, Task Force on Risk-Informed Inservice Inspection, dated August 2004. (report is easily retrievable with google search but not located in ADAMS) 2005 - European Framework Document for Risk-Informed In-Service Inspection, European Network on Inspection and Qualification (ENIQ), Task Group Risk (TGR). (couldnt find in google search or ADAMS but did find full citation: Chapman O J V, Gandossi L, Mengolini A, Simola K, Eyre T, and Walker A E (Eds.), European framework document for risk informed in-service inspection, ENIQ Report No. 23, JRC-Petten, EUR 21581/EN, 2005.

2005 - ASME Code Case N660 approved in Regulatory Guide 1.147, revision 14 for RI-categorizing pressure boundary components, dated August, 2002.

2006 - American Society of Mechanical Engineers, Case N-716, Alternative Piping Classification and Examination Requirements,Section XI, Division 1, April 19, 2006 2007 - Request for Alternative ANO2-R&R-004, Revision 1 Request to Use Risk-Informed Safety Classification and Treatment for Repair / Replacement Activities in Class 2 and 3 Moderate Energy Systems, dated April 17, 2007 (ADAMS Accession No. ML071150108) 2007 - Request for Alternative ANO2-R&R-004, Revision 1 Response to NRC Request for Additional Information, dated August 6, 2007 (ADAMS Accession No. ML072220160) 2009 - Arkansas Nuclear One, Unit 2-Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request to use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (TAC NO.

MD5250), dated April 22, 2009 (ADAMS Accession No. ML090930246) 2010 - EC-JRC/OECD-NEA Benchmark Study on Risk Informed Inservice Inspection Methodologies (RISMET), CSNI Integrity and Ageing Working Group (IAGE) dated November 2010 2010 - Using the EPRI Risk-informed ISI Methodology on Piping Systems in Forsmark 3, 2010:42, Swedish Radiation Safety Authority, SSM, dated December, 2010. (easily found with google search but not located in ADAMS) 2011 - COG Risk Informed In-service Inspection Pilot Study, COG-JP-4369-001, dated July 2011.

RA-23-0282 Enclosure Page 20 of 35 2012 - Vogtle Electric Generating Plant Pilot 10 CFR 50.69 License Amendment to Adopt 10 CFR 50.69, Risk-Request, dated August 31, 2012 (ADAMS Accession No. ML12248A035) 2013 - American Society of Mechanical Engineers, Case N-716-1, Alternative Piping Classification and Examination Requirements,Section XI, Division 1, January 27, 2013 2014 - Risk Informed In-service Inspection (RI-ISI) Methodology for CANDU Conventional Systems and Components, COG JP-4425, dated March 2014 2014 - ASME Code Case N716-1 approved in Regulatory Guide 1.147, revision 17, dated August, 2014 2014 - Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC NOS. ME9472 AND ME9473), dated December 17, 2014 (ADAMS Accession No. ML14237A034) 2015 - Periodic Inspection of CANDU nuclear power plant balance of plant systems and components, Canadian Standards Association, CSA N285.7-15, dated November, 2015 2018 - 10 CFR 50.69 Categorization Guidance Document, EPRI, Palo Alto, CA: 2018 3002012984 2019 - ENIQ Framework Document for Risk-Informed In-Service Inspection, Issue 2, ENIQ Report No. 51, NUGENIA Technical Area 8, dated March 2019. (easily retrievable with google search) 2019 - American Society of Mechanical Engineers, Case N-752, Risk Informed Categorization and Treatment for Risk-informed Repair / Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1, July 23, 2019 2020 - Relief Request Number EN-20-RR-001 - Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1, dated May 27, 2020 (ADAMS Accession No. ML20148M343) 2021 - American Society of Mechanical Engineers, Case N-752-1, Risk Informed Categorization and Treatment for Risk-informed Repair / Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1, April 12, 2021 2021 - Arkansas Nuclear One, Units 1 and 2 - Approval of Request for Alternative from Certain Requirements of the ASME Boiler and Pressure Vessel Code, dated May 19, 2021 (ADAMS Accession No. ML21118B039) 2021 - Single Source Document for Risk Informed Inservice Inspection Research: Summary of EPRI Research and Relevant Sources. EPRI, Palo Alto, CA: 2021. 3002021010.

2021 - Browns Ferry Nuclear Power Plant Units 1, 2 and 3 - Issuance of Amendment NOS.

317, 340 and 300 Regarding Adoption of Title 10 of the Code of Federal Regulations Section 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, (EPID l_2020-LLA-0162), dated December 15, July 27, 2021 (ADAMS Accession No. ML21173A177)

RA-23-0282 Enclosure Page 21 of 35 2022 - Guidance for Determination of Risk-Informed Safety Classification for Light Water Reactor Nuclear Facility Pressure Retaining Components, ASME N T B 2022, prepared by ASME Standards Technology, LLC (STLLC) and sponsored by The American Society of Mechanical Engineers (ASME) Nuclear Codes & Standards.

2022 - 10CFR50.55a Final Rule, ASME 2019-2020 Code Editions, approves use of Appendix R, Supplement 2 (TR-112657).

Figure 4. Excerpted from NRWG Report (Reference 11)

Let us rs: consider the case that e irn a.su a il re risk of e scope oonsidered. i s used to de*ennin is a RR re d.

• ishes be~n h "

say RR

= 1.005 o r

• dependent of each o nge. and co control ed n

e ng c r acteristi<:s and possible prob

+

The RR'

="-'-'="--'::<.:...-'-"'"-'-'-'~ included -

luation. By adding

• ems.

. at is.

tance full other i e m s can be changed n

agnitude of e effect depend on e in en e m en I

e increase in ris is la the ina-ease is small. the effect may be negligi 2.

law safety-cance o the u nd degradation m

• class 1 pi stem s in one subj not because of si:stan* m ateria l.

rel.a

• e risk pro les coul 3.

If ge con ns *.ems teep s lope. h " h ind ris or a risk pro e be concentrated at ery fe r risk profiles. hig s tribu ed :o more

• The erm

• order ature ooukf replacem us i :rod ta paradox 4. Assu or erro the i res degr.

and es chang fro put 63 a lso lead to a a *on e limi risk

• or change of opera

• s more h igh-safety-g item is d a d is i oenti e to lysis.

a o rstate o r unders'tate

• ms. can distort pa lar. undue oonservatisms in PSA. aJOocacon of o' failu re probabir *es can oversta e oen sa ty-sign* ca 5.

be e ff.

ely inspected by OT o r can

• be inspected by inspection.

Ids o r *

"b ra fatigue is the damage mecha ism.

y cant at the expense o' o ers at u ld be e r des e

RA-23-0282 Enclosure Page 22 of 35 Figure 5. Excerpted from RISMET Report (Reference 12)

Takeaway is that the CCDP approach identifies many RCS segments as medium or high safety significance while the RRW approach identifies many RCS segments as low or medium safety significance and few if any inspections required.

Table 19 Rm1kiug of segmeuts aud uumber of inspectious iu the Reactor Coolnut System.

313 ASME SKIFS RRW based W-S full l----'-.C.C.C'---+-+--+-+--l---+-+-- -+-+--+-+--+---+-+---I W-S4 W4EP W-S full EP W-S4 EP EPRIR4 CCN-716

RA-23-0282 Enclosure Page 23 of 35 Figure 6. Excerpt from BFN3 RI-ISI Template Submittal (Reference 13)

Takeaway is that the RRW approach identifies that over 90% of the piping as low safety significant while the CCDP based approach limits low safety significant to 35% of the piping primarily due to defense in depth considerations.

Table 3.7-3 SEGMENT CATEGORIZATION system

1. Segs Consequence category Risi< category High Medium Low Hl9'>

Mtdfvm LOW CCOP CCDP CCOP RRW RRW RRW

>1E-004

>1E-06,

<1E-06

~1.00$

?, 1.00*1,

<1.001

<1E-04

<1.005 001 MS 56 4

38 14 4

52 002 cow 38 26 10 36 003 FW 46 10 31 10 35 023 RHR.SW **

12 33 45 024 RCW 20 18 2

20 027 ccw 3

2 1

3 083 SLC s

5 5

067 EECW 28 7

17 28 068 RECIRC 16 16 3

4 059 RWCU 19 2

2 15 1

18 070 RBCCW 17 17 17 071 RCIC 12 3

12 073 HPCI 11 2

9 11 074 RHR 31 8

18 3

28 075 cs 15 3

8 4

2 13 078 FPC 1

1 1

085 CRO 31 7

24 31 total::

392 47 207 138 29 3

380

RA-23-0282 Enclosure Page 24 of 35 Figure 7. Excerpt from BVN3 RI-ISI Template Submittal (Reference 14)

Takeaway is that the RRW approach identifies that almost 90% of the piping as low safety significant and 75 % of the RCS as low safety significant.

Table 3.7-1 Summary of Risk Evaluation and Expert Panel Categorization Results System Number of Number of Number of Number of Number of Total segments segments segments segments segments number of with any with any with all with any with all segments RRW?.1005 RRW RRW<

RRW RRW<

selected for between 1.001 between 1.001 inspection

< 1.005and 1.005 and selected (High Safety

?.1 001 1.001 for Significant placed in inspection Segments)

HSS BO 0

0 27 0

27 27 (0)

CH 26 18 105 7

1 28 (28)

Cl 0

3 148 0

0 0 (0)

DV 0

0 7

0 0

0 (0)

FW 0

0 21 0

21 21 (0)

HY 0

0 32 0

0 0 (0)

MS 0

0 48 0

48 48 (8)

OS 0

2 46 0

3 3 (3)

RC 14 16 51 6

0 20 (20)

RH 0

1 37 1

18 19 (19)

RS 10 2

25 0

0 10 (10)

SI 34 42 78 7

1 30 (30) ss 0

0 44 0

0 0 (0)

TOTAL 84 84 669 21 119 206(118)

RA-23-0282 Enclosure Page 25 of 35 Table 1. Excerpt from WCAP-14572 Tables 3.5-2 and 3.5-3 (Reference 10)

System Small Full break break AFW - Auxiliary Feedwater3 8.80E-03 8.80E-04 CC - Component Cooling 8.20E-02 3.43E-02 CH - Chemical & Volume Control 1.22E-03 6.45E-07 CN -Condensate 3.60E-01 3.60E-02 Emergency Core Cooling 6.40E-08 1.50E-15 Feedwater 1.10E-03 3.50E-11 Reactor Coolant 1.30E-06 1.20E-12 High Pressure Safety Injection 4.10E-08 8.10E-13 Low Pressure Safety Injection 2.50E-08 9.20E-12 Service Water System.

1.70E-03 3.70E-13

RA-23-0282 Enclosure Page 26 of 35 Table 2. Excerpt from the Beaver Valley RI-ISI Template Submittal (Reference 14)

Table3.4-1 Faik.-e Probabilitv Estimates (without ISU S)'61.em Domha'II: Po:~

Degracl3:10n Farure Prot1.1011ny R3nge at ~o Ye.ar6 W1l No ISi eomments Med'llr!lonX&)/COmtllna:101'1$(&)

sma111ea 01531:i'.lng le.Jk (Cl'/<11&31:i'.lng leak ra:e)"

E1'061on.1eorro&ron, ThermJII Fa:lgUe-6.61E 4..9JE-05 SYS 1.70E*09

• 9.71E-D6 System Ii hCIUCle<'.I ti O'le CllnfY! FIOW'

~tea COrro&IOn AUgmentea 1n.......-*on......... ram.

CH TMnnoii f a:~

3..46E 2.17E-04 StOCA 1.4.2E*09

• 5JJ6E-05 SYS 9.03E*11

• 7.30E-05 ThMnoii & Vl)ratory FaGgu=

6.38E 5..SOE-03 StOCA 1.76E*OS

• 3.ESE-03 v 10ra.:on OOOJF6 near !he ort."!Oe& In vartou, SYS 6.99E*09

• 3.ESE-03 IOCa:JOni INOOs,n,l ~e sy&lem..

OOWnstream or tl'lE ctiargrig pufl1)& ana 1e1<<1wn ne.at e.xCIUl'lgef are tl'lE altJC<il onnce IOcatron&.

v 10ra.:on ooo.n on &m.MI orancn connection& near 0 1.171-Cl E1'061on.1eorrosron. Tnenna11 F a:gue 1.IOE 4.13E-07 SYS 1.32E*10

• 5.61E-09 MiCl'OOIOlogJC.31 mecna,nsm& were iaenunea 1n me-n\\-er water poruon& or ::ne plp(ng Hn.e&.

ThMnoiiFa:,..,..

1.09E 4.0SE-05 SYS 2.85E* 11

• U2E-05 0V TMnnoiif d.~

UOE*7-1.97E*S SYS 1.3'E* 10

• 4JJ6E-07 FW E1'061on.1eorro&ron, ~

Fa:lgue 1.52E t..&!E-06 SYS 1.18E* 12

• 1.76E-07 System Ii hCIUCle<'.I In Ile Cllnf'll FIOW' A<<E:eratea COrro&IOn AUgmentea 1.............. on......... ram.

HY TMnnoiiFd.~

267E 1.0SE-05 SYS 2.67E-07

• t.Oec-05 Gas segmem, wnere 1n a11 case& a &ma:1 iea WOl.f(I d:&atlle ::ne &y&li'!TI fl.l'lc:Gon.

MS E1'0610f\\lCOrJo&ron. ThermJII Fa:igue 279E-07

• 4..63E

• 04 SYS 2.SIE*I O

• 5.63E-05 Sy.&li'!TI IS hCIUCle<'.I In O'le Clffeni FIOW'

~tea COrro&IOn AUgmentea JnU\\IW""on,........ ram.

OS TMnnoiif d.~

5.AOE 2.61E-04 SYS 5.95E* 11

• 3JJ6E-05

\\lloratronal FallgUe 4.33E 3.21E-04 SYS 8.05E*09

• U3E-04 Vlora.:on ooo.n on &m.MI orancn connec:!lon& near "I.IT!""'.

Table 3.4-1 Faik.-e Probabilitv Estima1es twithoul ISi\\

Sy&li'!TI Domhant Po:err.'all DegracLT.IOn Fa1ure Prot1.1011ny Range at ~o Ye.ar6 w:,i No 1SJ comment&

Med'lll'llim{&)/COmtlln.a:101'1$(&)

smai11ea 01531:i'.lng le.Jk

,....,(11&31)'.IM leak ra:e,*

RC StMi COITT161on.'Cfaclrng, ThermJII 4.97E*10- t..88E-05 LLOCA 1.53E*12

• 3.67E-D6 FallgU<

MLOCA 2.0!E* 12

• 3.67E-D6 StOCA 1.02E*11

• 4.n E-D6 SYS 1.32E*11

• 9.33E-D6 StM& COITT161on,cracsrng. TMrmaJ 2.03E t.2SE-05 LLOCA SASE* 10

• 8.72E-D6 1n.aU&,y t11S:ory IGenmi.e, lharnal &'ll~ng Faogue, \\l!Dra:IOl'IJI F a:gue, MLOCA 5.9-.!E* 10

• 8.9SE-D6 or stra:l!leatron OCCU1$ tn th= pre~

Str'lpl~tffleallon StOCA 7.37E*10* 8.93E-D6

&l.l'ge 1rie.. T elTl)eralln monr.omg IS on SYS 1.11E*09

• 1.0I E-05 g<<ng ou; no e-Mence or fallgue- <<

&truc:tura1 concern& l'IJ*1e-oeen noti<I.

Water Hammel U 3E U 3E-07 SYS J.7JE* 10

• 4.4.!E*I O flli RH r nenna11Fa:g.ie 1.14E 8--70E-05 LLOCA 6.00E* 10

• 4.00E-07 MLOCA 6.06E* 10

• 4.IOE-07 StOCA 6.27E* 10

• 4.IOE-07 SYS 4.J3E* 16

• 7.22E-05 Tl'IEfmal Fa:g.,E. Vl:lratlon.11 Fatigue 3.0I E 3.61E-03 SYS 5.0SE* 10

• 2.SOE-03 viora:on ooo.n on &m.MI oran.cn connection& near 0 1.111m.

RS Tnemlalf d.~

1.07E-OS* t.53E~

SYS 6.n E-11

• 8.34E-05 Tl'lef'mal Fa:g.,E. Vl:lratlon.11 Fatigue 2.9'2E-OS* t.20E~

SYS 4.7~E-10

• 4.0:E-05 viora:on ooo.n on &m.MI oran.cn connection& near 01.1711:1$.

SI Tnemlal Fa:~

8.01E*1 t - 3.35E-04 LLOCA 2.09E-D8

• U SE-05 Tne-potentlail for tnennail &:rli:mg or MLOCA 6.2SE* 10

• 2.00E-05

&!ramtcatlon E-)!iti on &m~ orancn 1ries StOCA 2.63E*10* 1.01E-04 coma::rmg c:ne<< ~,

c:onneetea,o ::ne SYS 3.65E* 13

• 3.34E-05 malnlO"".

ss Tnemlal f a:*~

7.8-9E J.29E-05 SYS 3.S2E*10

• 2.20E-05 NO:E-.

• Ol&Jt>lhg !eat rate - LLOCA. MLOCA. SLOCA. ana SYS (&:,stem Cliablhg !eat).

RA-23-0282 Enclosure Page 27 of 35 RAI No. 2 Please provide the information presented during the audit about the example of the possible interaction effects if a valve with an active high safety-significant function and passive low safety significant function were to have the bonnet replaced using the allowances of Code Case N-752. Please address aspects of quality and design processes. The information related to this RAI was discussed during the August 16, 2023, audit call, and provided via e-mail dated August 22, 2023 (ML23269A041).

Duke Energy Response to RAI No. 2 During audit and public meeting discussions, the staff discussed the possible interaction effects if a valve with an active HSS function and passive LSS function were to have the bonnet replaced using the allowances of ASME Code Case N-752. An example as suggested by NRC staff of these potential interaction effects, would be if the valve bonnet were made of a different or lower quality material and had different thermal expansion properties such that the bonnet would interact with the moving parts of the valve (HSS function).

The code case specifically states:

Section 1420, item c: 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.

This requirement from the case is saying, for the example above (different bonnet material) that the licensee would evaluate the change in material to ensure the pressure boundary function would continue to support the design basis functional requirements of the system (i.e., the bonnet would retain the fluid and not allow the bonnet to impede the active function of the valve).

While the Relief Request implementation relaxes ASME Section XI and Quality Program Requirements, it does not alleviate Design Control process requirements. The Design Control program requires various levels of engineering evaluation based on the change implemented and whether that change is within the Bounding Technical Requirements, including evaluation of the changes per 10 CFR 50.59. For example, the 10CFR50.59 change control process does not allow changes if they:

(v) Create a possibility for an accident of a different type than any previously evaluated in the final safety analysis report; (vi) Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the final safety analysis report; Changes, such as changing the bonnet material, would be evaluated in the Duke Energy Design Control Process. The Duke Energy procurement process also requires Engineering to specify treatment (design, fabrication, testing, documentation, receipt, etc.) of the LSS SSCs to ensure reasonable confidence is maintained.

RA-23-0282 Enclosure Page 28 of 35 As noted in ASME Code Case N-752 Section 1420 the licensee has requirements to fulfill to ensure the alternate treatment is appropriate. These additional controls ensure continued capability and reliability of the design-basis function. Further, any conditions that may prevent an LSS SSC from performing its safety-related function under design-basis conditions will be identified and addressed in accordance with the licensees corrective action program (as discussed in Oconee Round 1 RAIs). In the example above, the valve bonnet must perform the pressure boundary function and must allow the active items of the valve to perform their function. Controls are established to ensure both of these functions are addressed for changes to the plant (via design control) and identification/resolution of undesirable conditions (via the corrective action program).

The licensee must continue to implement all special treatments applied to active components (e.g. In-Service Testing) and other special treatments to the pressure boundary components that are not affected by this code case (e.g. In-Service Inspection, License Renewal Aging Management, Flow Accelerated Corrosion, Erosion, Raw Water Program, Buried Pipe Program). For a valve there are extensive tests and design calculations that would explicitly test for degradation via robust testing and the design calculations also would explicitly look for such material property changes.

This relief request does not change the Oconee Technical Specifications and all Surveillance Requirements will continue to be performed with the specified Frequencies. Oconee SR 3.0.1 governs and provides usage rules for all Surveillance Requirements.

As noted in Oconees Relief Request and Round 1 RAIs, 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. In accordance with 10 CFR 50.54(a), licensees may make changes to a previously accepted quality assurance program description provided the bases of the approval are applicable to the licensees facility. Duke Energy does not intend to deviate from the QAPD wording in the NRC Safety Evaluation for ANOs QAPD Reduction in Commitment utilized to implement ASME Code Case N-752. The NRCs review of the ANO QAPD changes to implement ASME N-752 states [bolding added for emphasis]:

The NRC staff reviewed the licensees application, as supplemented, for the proposed change to the QAPM to allow the use of ASME Code Case N-752. The NRC staff has 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. This reasonable assurance is based on the licensees plans to continue using current processes and procedures, as supplemented, for the treatment of Class 2 and 3 LSS SSCs. In addition, when procuring Class 2 and 3 LSS SSC items, the licensee will specify supplemental procurement requirements and implement additional controls to ensure continued capability and reliability of the design-basis function. Therefore, the NRC staff concludes that the proposed Entergy QAPM change continues to provide an acceptable level of quality and safety.

In order to implement the same change to the Oconee QAPD without prior NRC approval, via the process outlined in 10 CFR 50.54a, Duke Energy must demonstrate the bases for the NRCs QAPD approval at ANO is applicable to Duke Energy. The same QAPD changes the NRC found to provide an acceptable level of quality and safety for ANO will also be implemented at Duke Energy (Oconee specifically).

RA-23-0282 Enclosure Page 29 of 35 RAI No. 3 Code Case N-752 allows for broad application of codes and standards, as well as Owner's Requirements in addressing repair/replacement and non-destructive examination activities in accordance with Section 5.2.E of the licensees proposed alternative.

Please provide the information presented during the audit as it relates to the use of Owners Requirements to meet an acceptable level of quality and safety under 10 CFR 50.55a(z)(1) for the proposed alternative. The information related to this RAI was discussed during the September 20, 2023, audit call, and provided in an e-mail dated September 26, 2023 (ML23270B836).

Duke Energy Response to RAI No. 3 NRC Audit Question from September 15, 2023 that was discussed during the September 20, 2023 audit call is as follows:

In its submittal dated July 27, 2022, as supplemented by March 9, 2023, the licensee states that the use of ASME Code Case N-752 is requested with no exceptions or deviations. The NRC has not approved ASME Code Case N-752 generically. As discussed during the NRC audit and related discussions concerning possible information needs to make a regulatory decision, the NRC staff requests the licensee to clarify:

What code/standard would apply for alternative treatment when implementing ASME Code Case N-752 for repair/replacement of passive components? The current provision in N-752 allows for broad application of "Owner's Requirements" in addressing repair/replacement and non-destructive examination activities in accordance with Section 5.2.E Paragraphs 4 through 9, and 16 of the licensees submittal. Some confirmation or clarification is needed regarding the applicable code/standard to evaluate uncertainty in the change in safety margins. For NRC to reach a finding of acceptable quality and safety additional confirmation is needed regarding the applicable code/standard. An example of such codes/standards include the ASME Boiler and Pressure Vessel Code, ASME B31.1, Power Piping Standard and/or ASME Post Construction Committee 2 (PCC-2) Repair of Pressure Equipment and Piping), which the NRC staff have found to be acceptable in similar application for repair/replacement and NDE for passive components. The NRC recognizes that specifying the code/standard may be an exception/deviation from N-752, nevertheless, the staff may issue a request for confirmation that the code/standard in the analysis of record still applies or request for additional information to clarify if another code/standard is proposed.

Duke Response to September 15, 2023 NRC Audit Question:

Safety margins for LSS items, in accordance with ASME Code Case N-752, are maintained by the alternative treatment process ensuring, with reasonable confidence, the items remain capable of performing its safety-related function under design-basis conditions along with the other programs in place at Oconee to monitor, inspection, operate, and maintain installed equipment.

The treatment process is described in Oconee Relief Request (ML22208A031), Section 5.2.E and is outlined below.

RA-23-0282 Enclosure Page 30 of 35 ASME Case N-752, Para. -1420(e) states, Items used for repair/replacements shall meet the original Construction Code to which the original item was constructed. Alternatively, items used for repair/replacement activities shall meet the technical requirements of a nationally recognized code, standard, or specification applicable to that item (e.g. ASME, ANSI, AWS, AISC, AWWA, API 650, API 620, MSS SPs, or TEMA), as permitted by the licensing basis.

There are a wide variety of potential uses for this code case at Oconee. To that end, Duke Energy will evaluate activities in light of the variety of solutions depending on the issue, availability of products/solutions, and evaluation of technically acceptable solutions (i.e.

reasonable confidence evaluation). Example alternatives that could be used, and would likely be used in many cases, include but are not limited to the following:

Alternative Piping Codes: ASME B31.1 instead of ASME Section III; ASTM materials specifications instead of ASME Section II Alternative Valve Codes: ASME B31.1, ANSI B16.34 instead of ASME Section III Alternative Vessel Codes: ASME Section VIII instead of ASME Section III Alternative Tank Codes: API-620, API-650, AWS B96.1 Additionally, ASME Case N-752, Para. -1420(g) requires that performance of repair/replacements including NDE shall comply with the Construction Code, Post-Construction Code, or Standard selected for the repair/replacement activity. Examples alternatives that could be used, and would likely be used in many cases, include but are not limited to the following:

Alternative Construction Codes: ASME B31.1, ASME Section VIII instead of ASME Section III Alternative Post-Construction Codes: ASME PCC-2, API-653 instead of ASME Section III Alternative Standards: AWS D1.1, D1.3, D.16 Further, ASME Code Case N-752, Para. -1420(b) requires, even when using alternative Construction Codes, Post-Construction Codes, Standards, and Specifications, an Owner to comply with the fracture toughness requirements of the original Construction Code and Owners Requirements.

ASME Code Case N-752, para. -1420 specifies that the Owner is responsible for confirming with reasonable confidence that each LSS items remains capable of performing its safety-related functions under design-basis conditions when defining requirements for design, procurement, installation, etc. on LSS items. Meaning, Owners cannot select an erroneous Code or Standard for performing repair/replacement activities on LSS items. In fact, many of the Codes that will be used for LSS items are those that are already being used by non-Section III plants. ASME Code Case N-752 provides much more guidance and includes limitations on alternative treatments that do not exist in related applications like 10 CFR 50.69. The Statements of Consideration (SoC) for 10 CFR 50.69 evaluates/documents the agencys position on application of Voluntary Consensus standards as a part of Alternative Treatment.

During the 50.69 rulemaking process, ASME and other stakeholders did not recommend adding

RA-23-0282 Enclosure Page 31 of 35 a provision onto the 50.69 rule itself because the SoC and regulation provide adequate guidance for treatment. At the time, the NRC did note a wide variation existed in industrial practices and there may be certain industrial practices that may not be sufficient to satisfy the treatment requirements for LSS components. To address these concerns, the Commission clarified the rule requirements, and ASME Code Case N-752 adopted these clarifications (see Paragraph -1420), to indicate that the treatment of RISC-3 / LSS SSCs must be consistent with the categorization process. While one way to achieve this consistency could be the application of specified consensus standards, it was decided in 10 CFR 50.69 rulemaking as well as in the development of ASME Code Case N-752, that the Owners requirement to confirm, with reasonable confidence, that RISC-3 or LSS SSCs remain capable of performing their safety-related functions under design basis conditions, was sufficient for these LSS components.

The 10 CFR 50.69 and ASME Code Case N-752 categorization processes use a 1.0 failure probability which provides a conservative risk-informed categorization result and also minimizes uncertainty. The categorization process assumes failure in all cases which shows, regardless of what treatment code or standard is used, the categorization process bounds the change in failure frequency arising from changes in treatment. Nevertheless, as stated above, 10 CFR 50.69 and ASME Code Case N-752 (e.g. -1420) requires that the Owner ensure with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design-basis conditions.

Implementation of ASME Code Case N-752 only allows the licensee flexibility on ASME Section XI and QA requirements. Other programs and process remain in place such as design control, 10 CFR 50.59 change control process, supply chain / procurement processes, corrective action

/ problem identification and resolution, testing and monitoring programs (e.g. RI-ISI, IST, License Renewal Aging Management, Flow Accelerated Corrosion, Erosion, Raw Water Program, Buried Pipe Program, etc.), and Technical Specifications (including surveillances).

These programs allow the licensee to monitor the condition of components, identify degradation, and correct the degradation in a timely manner.

As noted in Oconee Relief Request (ML22208A031), Section 5.2.E, Item 14, Duke Energy intends to utilize the provisions of 10 CFR 50.54(a)(3)(ii) to make changes to the QAPD without prior NRC approval. As noted in this regulation, the change to the QA program can only be completed if the bases of the NRC approval are appliable to the licensees facility. This would include additional requirements, like those noted in Entergys RAIs dated April 30, 2021 (ML21120A326), for supplemental procurement requirements and controls for ASME Code Case N-752 LSS items.

Implementation of ASME Code Case N-752 in its entirety ensures safety margins are reasonably maintained by meeting all aspects of treatment outlined in the request and ensuring with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design-basis conditions.

Duke Energy Owners Requirements Discussion Provided in September 26, 2023 e-mail:

Owner's Requirements is a defined term with specific meaning in Section XI that distinguishes Owners Requirements from Code requirements, establishing two separate sets of requirements. This is a Section XI convention that exists in the base code editions/addenda and was implemented in Code Case N-752 and the Oconee submittal to retain consistency with code editions/addenda in use in the industry.

RA-23-0282 Enclosure Page 32 of 35 In other words, the Design Specification requirements for an item are comprised of code requirements, regulatory requirements (e.g. conditions applicable to a construction code), and Owner's Requirements.

Duke submittal paragraph 5.2.E.4, patterned after Code Case N-752, -1420(d), states:

Items used for repair/replacement activities shall meet the Owners Requirements or revised Owners Requirements as permitted by the licensing basis.

Duke submittal paragraph 5.2.E.5, patterned after Code Case N-752, -1420(e), states:

Items used for repair/replacement activities shall meet the Construction Code to which the original item was constructed. Alternatively, items used for repair/replacement activities shall meet the technical requirements of a nationally recognized code, standard, or specification applicable to that item as permitted by the licensing basis.

Application:

Paragraph 5.2.E.4 establishes the Owners Requirements for items to be used for repair/replacement under Code Case N-752. Since Owner's Requirements is given specific meaning distinct from Code requirements, paragraph 5.2.E.5 is needed to establish code requirements for an item to be used for replacement under Code Case N-752. The two sets of requirements comprise a full set of requirements for an item, Owners and Code Requirements. Submittal paragraphs 5.2.E.4 and 5.2.E.5 are parallels to the 2017 edition of ASME Section XI, IWA-4221, Construction Code and Owners Requirements which describes use of Owners and Code requirements. These two paragraphs are consistent with current Section XI code requirements.

Paragraph 5.2.E.4 is met by either meeting the original Owner's Requirements for the item or revised Owner's Requirements. This paragraph does not discuss use of Owner's Requirements in lieu of Code requirements and does not grant any permissions in that regard. Revised Owner's Requirements must be evaluated as required by 5.2.E.3.

Paragraph 5.2.E.5 requires that replacement items meet the original Construction Code or, alternatively, technical requirements of a nationally recognized code, standard, or specification applicable to that item as permitted by the licensing basis. Thus, Owner's Requirements cannot wholly replace Construction Code requirements because items must meet one of these two 5.2.E.5 requirements (original construction code or alternate code).

The continued use of distinct Owners and Code requirements in Code Case N-752 and the submittal has been similarly established in code edition/addenda, and the specific requirements provided for each to prevent indiscriminate substitution of Owners Requirements for Code requirements.

Code Case N-752 does provide two specific owner permissions and similar permissions are requested in the submittal in paragraphs 5.2.E.7 and 5.2.E.8 which will be discussed below.

Duke submittal paragraph 5.2.E.6, patterned after Code Case N-752, -1420(f), states:

RA-23-0282 Enclosure Page 33 of 35 The repair methods of nationally recognized post-construction codes and standards (e.g., PCC-2, API-653) applicable to the item may be used.

Application:

This paragraph allows the use of nationally recognized post-construction codes and standards applicable to the item. 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.

Duke submittal paragraph 5.2.E.7, patterned after Code Case N-752, -1420(g), states:

Performance of repair/replacement activities, and associated nondestructive examination (NDE), shall be in accordance with the Owners Requirements and, as applicable, the Construction Code, or postconstruction code or standard, selected for the repair/replacement activity. Alternative examination methods may be used as approved by the Owner. NDE personnel may be qualified in accordance with IWA-2300 in lieu of the Construction Code.

Duke submittal paragraph 5.2.E.8, patterned after Code Case N-752, -1420(h), states:

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.

Application:

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 activities (e.g. welding, brazing, defect removal, procurement, design, fabrication, installation, examination and pressure testing of items) will follow a nationally recognized Construction Code, or postconstruction code or standard similar to the requirement in submittal paragraph 5.2.E.5.

It also allows alternate NDE examination methods which are discussed below along with paragraph 5.2.E.8.

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. Paragraph 5.2.E.7 allows for alternative NDE methods and 5.2.E.8 allows for pressure testing requirements to be determined by the owner.

However, these permissions are also subject to paragraph 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.

RA-23-0282 Enclosure Page 34 of 35 Duke submittal paragraph 5.2.E.9, patterned after Code Case N-752, -1420(i), states in part:

Baseline examination (e.g., preservice examination) of the items affected by the repair/replacement activity, if required, shall be performed in accordance with requirements of the applicable program(s) specifying periodic inspection of items.

Application:

In paragraph 5.2.E.9, the if required statement pertains to the scoping of the preservice ISI or IST examinations. The ASME Code Case N-752 LSS components may, or may not, be within the scope of those programs. If the ASME Code Case N-752 LSS components are not within the scope of those programs, then the examinations are not required. If the ASME Code Case N-752 LSS components are within the scope of those programs, then the examinations are determined by the program requirements.

Duke submittal paragraph 5.2.E.16 states:

As permitted by Code Case N-752, Duke Energy intends to implement the exemption on IWA-4000 applicable to repair/replacement activities. Article IWA-4000 of the ASME Section XI Code specifies administrative, technical, and programmatic requirements for performing repair/replacement activities on pressure-retaining items and their supports. As specified in IWA-4110(b), repair/replacement activities "include welding, brazing, defect removal, metal removal by thermal means, rerating, and removing, adding, and modifying items or systems.

These requirements are applicable to procurement, design, fabrication, installation, examination, and pressure testing of items within the scope of this Division". In lieu of these IWA-4000 requirements, Duke Energy will perform repair/replacement activities on LSS items in accordance with an Owner-defined program that complies with paragraph -1420 of Code Case N-752. The Duke Energy program will utilize existing Duke Energy processes such as those applicable to procurement, design, re-rating, fabrication, installation, modifications, welding, defect removal, metal removal by thermal processes and supplement those process requirements as necessary to comply with Code Case N-752. Duke Energy believes this program will ensure, with reasonable confidence, that LSS items remain capable of performing their safety-related functions under design basis conditions. Finally, the exemption of IWA-4000 requirements and the alternative use of Owner-defined treatment requirements for LSS items is consistent with the NRCs position on risk-informed programs as specified in 10 CFR 50.69(b)(1)(v) and (d)(2).

Application:

This paragraph is stating Duke will use existing infrastructure (e.g. Welding Program, Design Control, Procurement, procedures and practices, etc.) to implement Code Case N-752. Specifically, how the activities listed in paragraph 5.2.E.16 would be completed are addressed in items 5.1.E 1-17.

==

Conclusion:==

Select repair/replacement tasks (e.g. NDE methods, pressure testing) allow the Owner to establish requirements for the activity. Code Case N-752 requires that repair / replacement activities on LSS items follow a nationally recognized Construction Code, or post-construction code or standard, with the possible exception of Owner defined NDE methods and pressure testing. These requirements confirm, with reasonable confidence, that each LSS item will

RA-23-0282 Enclosure Page 35 of 35 remain capable of performing its safety-related functions under design-basis conditions.

Furthermore, the Owner is required to ensure the structural integrity and leak tightness of the system are sufficient to support the design basis functional requirements of the system.

Condition monitoring programs (e.g. Technical Specifications, RI-ISI, License Renewal Aging Management, etc.) remain in effect to provide continued assurance that unknown degradation of these pressure retaining items is not occurring. Further, these condition monitoring programs monitor for operational leakage, which has the effect of ensuring Technical Specification structures, systems, and components remain operable and able to perform their safety function.

Finally, LSS categorized components that are classified as ASME BPV Code Class 2 and 3, and are required to be operable by plant Technical Specifications, will be treated in accordance with the licensees approved relief request requirements (i.e., in accordance with Code Case N-752). Even though the categorization process assumes a failure probability of 1.0 and shows that failure of these LSS SSCs will have a negligible impact on public health and safety, the relief request requires that, with reasonable confidence, LSS SSCs remain capable of performing their safety related functions under design basis conditions.