ML24352A481

Final Request for Additional Information - EPRI Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components

ML24352A481
Person / Time
Site:	Electric Power Research Institute
Issue date:	01/13/2025
From:	Lois James Licensing Processes Branch
To:	Ferrante F Electric Power Research Institute
References
EPRI TR 3002025288, EPID L-2023-TOP-0031 pre-app, EPID L-2023-TOP-0045 pre-fee, EPID L-2023-NTR-0008 post-fee
Download: ML24352A481 (9)

Text

EPRI REPORT 3002025288, ENHANCED RISK-INFORMED CATEGORIZATION METHODOLOGY FOR PRESSURE BOUNDARY COMPONENTS REQUESTS FOR ADDITIONAL INFORMATION (RAIS)

Background:

By letter dated August 17, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23234A266), as supplemented on November 30, 2023 (ADAMS Accession No. ML23334A210), and June 14, 2024 (ADAMS Accession No. ML24180A016), Electric Power Research Institute (EPRI) submitted EPRI Technical Report (TR) 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components, dated June 2023, to the U.S. Nuclear Regulatory Commission (NRC) for review and approval. EPRI TR 3002025288 presents an enhanced methodology for categorizing pressure boundary components in support of Title 10 of the Code of Federal Regulation (10 CFR) Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, applications.

By letter dated July 11, 2024 (ADAMS Accession No. ML23352A054), the NRC staff accepted EPRI TR 3002025288 for review.

Regulatory Basis: Section 50.69 of 10 CFR provides an alternative approach for establishing requirements for treatment of SSCs for nuclear power reactors using a risk-informed method of categorizing structures, systems and components (SSCs) according to their safety significance.

Specifically, for SSCs categorized as low safety significance (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs determined to be of high safety significance (HSS), requirements may not be changed. The corresponding statement of considerations (SoC) for the rulemaking are in the Federal Register notice published on November 22, 2004 (69 FR 68008).

Paragraph 50.69(b)(3) of 10 CFR states that the Commission will approve a licensees implementation of this section by issuance of a license amendment if the Commission determines that the categorization process satisfies the requirements of 10 CFR 50.69(c).

Paragraph 50.69(c)(1) of 10 CFR states, in part:

SSCs must be categorized as Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, or RISC-4 SSCs using a categorization process that determines if an SSC performs one or more safety significant functions and identifies those functions. The process must:

(ii)

Determine SSC functional importance using an integrated, systematic process for addressing initiating events (internal and external), SSCs, and plant operating modes, including those not modeled in the plant-specific probabilistic risk assessment (PRA). The functions to be identified and considered include design bases functions and functions credited for mitigation and prevention of severe accidents. All aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.

(iii)

Maintain defense-in-depth.

(iv)

Include evaluations that provide reasonable confidence that for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in core damage frequency (CDF) and large early release frequency (LERF) resulting from changes in treatment permitted by implementation of §§ 50.69(b)(1) and (d)(2) are small.

(v)

Be performed for entire systems and structures, not for selected components within a system or structure.

Requests for Additional Information (RAIs)

RAI 01 - Accounting for High Consequence Scenarios Background/Issue:

EPRI report criteria 11-13 propose to categorize SSCs as LSS if their individual contribution to CDF is less than 10-6 per year, or if the SSC contribution to LERF is less than 10-7 per year without any consideration of consequences. Further, for CDF contribution between 10-6 per year and 10-8 per year or LERF contribution between 10-7 and 10-9 per year, a sliding scale of consequence consideration of conditional core damage probability (CCDP) or conditional large early release probability (CLERP) of 1.0 or greater than 0.1 or 0.01 is introduced for HSS categorization.

Based on a review of the documents in the audit, the NRC staff found SSCs with a CCDP of greater than 10-4 and a CDF contribution of less than 10-6 per year. Some internal flooding analyses have identified areas with CCDPs greater than 10-3 but would be LSS using the proposed 14 criteria. RISC-3 LSS SSCs would not be covered by American Society of Mechanical Engineers (ASME) Code or 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, requirements and can be repaired or replaced with uncodified nonmetallic repairs with no significant operating experience or lower-quality materials with unknown failure probabilities.

The staff notes that prior approved precedents (such as: EPRI TR-112657, Revised Risk-Informed Inservice Inspection Evaluation Procedure" (ADAMS Accession No. ML013470102);

ASME Code Case N-660, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement ActivitiesSection XI, Division 1," ASME Code Case N-752, "Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1," ANO2-R&R-004, Rev. 1, "ANO-2 Risk-informed Repair and Replacement Methodology" (ADAMS Accession No. ML071150108) currently approved by the NRC for categorization of passive components) considered failure scenarios with a CCDP greater than 10-4, or CLERP greater than 10-5, as high consequence scenarios, resulting in HSS categorization for the corresponding SSCs. Additionally, these approved precedents were consequence-based evaluations, where only the consequences of a postulated passive component failure were evaluated, and the failure frequencies or contribution to CDF/LERF were not taken into account. In contrast to prior approved precedents, the proposed EPRI methodology uses products of CDF (LERF) and CCDP (CLERP) as a comparison to CDF (LERF) which can contain a pipe rupture frequency as low as 10-8 per year (10-9 per year for LERF and CLERP) coupled with a consequence (CCDP or CLERP) as high as 1.0 and still be categorized as LSS. The staff finds the change in CCDP/CLERP thresholds and the introduction of CDF/LERF contributions are insufficiently justified to categorically conclude that those SSCs would be LSS subject to alternative treatments.

Requests:

a. The staff has identified the issue above regarding potentially not adequately addressing high consequence failures, specifically consequence failures with CCDP greater than 10-4 or CLERP greater than 10-5. Discuss how EPRI intends to modify the TR to address these areas. As discussed during the audit, please provide consideration of the following two options:

Provide a description and justification of how high consequence SSCs with CCDP greater than 10-4 or CLERP greater than 10-5 are addressed.

As an alternative to first item above, provide a clearly defined minimum set of requirements such as the repair methods of nationally recognized postconstruction codes and standards (e.g., ASME B31.1, ASME PCC-2) for SSCs with a CCDP of 10-4 and CLERP of 10-5 or higher.

RAI 02 - SSC Categorization as a Single Plant Unit Background/Issue: 10 CFR 50.69(c)(1)(v) requires that the 10 CFR 50.69 categorization process be performed for entire systems and structures, not for selected components within a system or structure. The final rules SoC explain that This required scope ensures that all safety functions associated with a system or structure are properly identified and evaluated when determining the safety significance of individual components within a system or structure and that the entire set of components that comprise a system or structure are considered and addressed.

EPRI TR 3002025288 Section 4.4, Alternative Treatment Requirements Under 10 CFR 50.69(d)(2), states that this enhanced methodology defines the pressure boundary function of each individual plant unit as a system for 10 CFR 50.69 categorization and alternative treatment purposes. Consistent with 10 CFR 50.69 rule language and several citations in the final rules SoC, the system boundaries for the pressure boundary function are limited to pressure retention. Therefore, there will be no other important functions that would escape categorization and appropriate assignment of safety significance. As covered in the Statements of Condition, this ensures that all safety functions in the selected system are properly identified and categorized regarding their safety significance.

Further, Table 7 of the TR, Comparison to 10 CFR 50.69(c)(1) states that the enhanced methodology requires categorization of all systems providing a pressure boundary function.

The statements in the TR appear to imply that all the pressure-retaining components in the plant are considered as one system and that only the pressure-retaining function will be used to define the system. The staff does not find the TR provides sufficient explanation on how the proposed passive categorization will be implemented in the overall 50.69 categorization and did not find sufficient justification to support the statement that all safety functions in the selected system are properly identified and categorized regarding their safety significance.

Figure 1 of the TR, Categorization process overview, depicts the overall 50.69 categorization process as intended to be implemented, and shows that the passive categorization is performed in parallel to the other aspects of the categorization, such as considerations based on PRA and other qualitative consideration. All these aspects are considered for the preliminary categorization step, per the guidance in Nuclear Energy Institute (NEI) 00-04 Section 7. The guidance in NEI 00-04 Section 7 states that SSCs that support multiple functions should be assigned the highest risk significant of any function that the SSC, or part thereof, supports.

Finally, the inputs from the preliminary categorization are provided to the integrated decision-making panel (IDP). Table 1 of the TR states that the IDP cannot change categorization from HSS to LSS for passive components.

Requests:

a.

Clarify what is meant by that statements that the methodology defines the pressure boundary function of each individual plant unit as a system. Describe how the EPRI methodology proposes to organize passive components in systems for the purpose of the passive component categorization. Describe how system functions are defined per Section 4 of NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline in the context of passive SSCs.

b. As indicated in Figure 1 of the TR, the passive categorization is one aspect of the systematic and integrated categorization process outlined in NEI 00-04. Describe further how the passive categorization will be executed part of the overall integrated categorization process. Describe how the guidance in NEI 00-04 Section 7 for preliminary categorization will be implemented for passive components. How will all aspects of the categorization process be considered for the preliminary categorization of SSCs (both active and passive) that will be provided to the IDP?

c. Describe IDPs role in addressing both the passive and active functions of SSCs.

Confirm the intent in TR Table 1 that IDP will not change HSS categorization of passive components.

d. Justify how the approach taken in EPRI TR 3002025288 for passive pressure boundary SSC categorization complies with 10 CFR 50.69(c)(1)(v) and the associated statements of considerations to ensure that all safety functions associated with a system or structure are properly identified and evaluated when determining the safety significance of individual components within a system or structure and that the entire set of components that comprise a system or structure are considered and addressed.

e. Can the proposed methodology create a situation where a component is only categorized for its passive function, but the associated active function is left uncategorized? If so, explain why this is acceptable. Also, if the proposed methodology can create situations where a single SSCs receives different categorization based on its active and passive functions, describe and justify such scenarios and the mechanisms on how that would occur. For each scenario, describe and justify how it is ensured that an active HSS function would not be impacted by the LSS designation of a passive SSC that supports that function. Describe the guidance and approach for resolving differences.

RAI 03 - Addressing Uncertainty and Other Events in Individual Assessments Background/Issue: Paragraph 50.69(c)(1)(i) states that the SSC categorization process must consider results and insights from the plant-specific PRA. This PRA must, at a minimum, model severe accident scenarios resulting from internal initiating events occurring at full power operation. The PRA must be of sufficient quality and level of detail to support the categorization process and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC. In response, Table 7 of TR 3002025288, Comparison to 10 CFR 50.69(c)(1), further states, As stated previously, the plant needs to have a robust internal events PRA, including IF [internal flooding], that addresses failure of all pressure boundary components (main steam line breaks, main feedwater line breaks, internal flooding events, interfacing system LOCA [loss of coolant accident], and so on).

Because this methodology is being used in support of 10 CFR 50.69 applications, the plant-specific PRA needs to be sufficient to support the license amendment request approval process, including consideration of PRA assumptions and sources of uncertainty.

Requests:

a.

Criteria 11-13 are the only criteria in the methodology that involves a direct use of the licensees PRA model-of-record. From Figure 3, CCDP versus CDF threshold and Figure 4, CLERP versus LERF threshold, it appears that each of the three criteria have hard risk thresholds. Explain how uncertainty is taken into account within the use of these thresholds to categorize a passive pressure-retaining component. Also, explain how the potential cumulative impact of changes is addressed.

b.

Discuss and justify how current risk thresholds for Criteria 11-13 take into account cases of lower initiating event frequencies coupled with higher failure consequences. Discuss how these higher failure consequences are considered.

c.

As a risk-informed process, discuss how the preservation of defense-in-depth and maintenance of safety margins are accounted for in using Criteria 11-13.

Also elaborate on the assessment of qualitative criteria and defense-in-depth for passive categorization, and if any additional guidance is required for the IDP when applying the methodology in EPRI TR 3002025288.

d.

In computing the CDF/LERF and CCDP/CLERP for Criteria 11-13, discuss how various embedded events in PRA models such as recovery actions (i.e., FLEX) and human reliability analyses are taken into account.

e.

For plants which have a high seismic contribution to pipe rupture, discuss how the results of various analyses (e.g., seismic PRA, Seismic Margins Analysis) are taken into account for Criteria 11-13. If these considerations are addressed qualitatively, please explain how they will be addressed. If these considerations are addressed solely by the IDP, explain how this is communicated to the IDP and what guidance is available for the IDP.

RAI Qualitative Considerations for Shutdown Operations and External Events Background/Issue: Section 2 of EPRI TR 3002025288 describes how the 10 CFR 50.69 categorization process is performed in accordance with NEI 00-04, Revision 0, as endorsed in Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, Revision 1.

Figure 1 of EPRI TR 3002025288, Categorization process overview shows passive categorization as a separate path for preliminary categorization of pressure-retaining components, prior to IDP review and final categorization. The NRC notes that the guidance in NEI 00-04 includes considerations for fire, seismic, and other external hazards, which may be assessed qualitatively, as well as additional qualitative criteria and requirements for assessment of defense-in-depth. (These are also shown in Figure 1.) The NRC notes that the NRC approved methodology for passive categorization in ANO2-R&R-004, Revision 1, also includes considerations for assessing shutdown operations, external events, and DID.

Requests:

a.

It is unclear how the methodology in EPRI TR 3002025288, whether independently or in conjunction with the guidance in NEI 00-04, requires the assessment of shutdown operations and external events, including external events that do not affect likelihood of pressure boundary failure but create demands that might cause pressure boundary failure and events (e.g., fires), for potential impact on the categorization of passive, pressure-retaining components.

Please explain how these considerations are addressed. If these considerations are addressed solely by the IDP, explain how this is communicated to the IDP and what guidance is available for the IDP.

RAI 05 - Plant Design Scope for Methodology Background/Issue: EPRI TR 3002025288 states that plant-specific 10 CFR 50.69 system categorization was verified for robustness through evaluation of several boiling water reactor (BWR) and pressurized water reactor (PWR) plants. This group of plants is described as consisting of multiple designs and included earlier-vintage and later-vintage designs. No basis is given for the generic applicability of the EPRI TR to other designs, including ones not yet designed. No limitation is presented in the report regarding which designs the report may be applied.

EPRI clarified, during the audit, that the methodology was not verified for designs such as the NuScale US600 or Westinghouse AP1000. The NRC staff needs this clarified on the docket.

Requests:

a.

If the methodology was not meant to include such designs, or other future PWR and BWR designs, provide revisions limiting the use of the methodology to the designs for which it was verified.

b.

If the methodology is meant to include designs other than those evaluated, clarify how this was verified and how it is controlled within the methodology.

RAI 06 - Clarification for Required Prerequisite Programs to the Methodology Background/Issue: EPRI TR 3002025288 includes, in Section 4.1, that robust program[s] for localized corrosion, flow accelerated corrosion (FAC), and erosion must be ensured before implementing the categorization in Section 4.2 of the methodology. The necessary quality and effectiveness of such programs is verified through, self-assessment, benchmarking, or peer review for localized corrosion; and reference to EPRI reports for FAC and erosion. The descriptions include optional language such as should.

During the audit EPRI provided an example application that relied on referencing individually identified EPRI guidance documents and an NEI bulletin. Further, it was clarified that applicants changing their programs may fall outside of the methodology, despite this not being explicitly controlled in the methodology. The NRC staff needs this clarified on the docket.

Requests:

a.

It is unclear within the methodology whether an applicant must meet the descriptions of the three programs or what alternatives would be acceptably similar. Clarify how this should be determined and whether optional elements of the descriptions (those including language like should) are genuinely optional.

b.

On what basis are the cited programs, or alternatives chosen by an applicant, determined to be sufficiently robust, and what would constitute an indication that these programs were insufficiently robust in implementation or due to future alterations?

c.

The methodology does not explicitly require that these programs continue after implementation of the methodology.

1.

How is this controlled in the methodology if these programs were discontinued or modified?

2.

How would an applicant referencing this methodology determine whether modifications supported a sufficiently robust program?

d.

The methodology references specific revisions of EPRI reports as necessary robust programs. Describe the process of how an applicant using the methodology will do if or when those references were updated?

e.

Would the 50.69 categorization need to be revisited if the referenced EPRI reports are revised and/or otherwise become insufficiently robust?

RAI 07 - Reference to Industry Guidance for Quantitative Assessment Background/Issue: EPRI TR 3002025288 includes, in Section 4.2 under Criteria 11-13, that users should rely on industry guidance for a number of risk impacts. It is unclear if NRC review and approval is being sought to generically accept use of unspecified industry guidance (examples are given but are not required) as being sufficient for regulatory review of performance of Criteria 13.

During the audit, EPRI stated that this could be clearer to refer to Prerequisite 4.1.1. The NRC staff needs this clarified on the docket.

Requests:

a.

Please confirm or clarify if this was the intent.

b.

Please clarify what industry guidance is meant and for which purpose it is to be used.

RAI 08 - Clarification of Reactor Coolant Boundary Categorization Background/Issue: EPRI TR 3002025288 Criteria 1 differentiates components based on whether the components can be isolated from the reactor coolant system by two valves in series. Table 3, HSS criteria: considerations, amends this to note that the piping between these two valves may be medium/low consequence. It is unclear how a valve whose function is dependent on a lower classification can retain a higher classification function as a matter of categorization.

During the audit EPRI stated that this could have been more clearly worded and provided a proposed revision. The NRC staff needs this clarified on the docket.

Requests:

a.

Submit the proposed revision, similar equivalent, or otherwise clarify why such is not needed.

RAI 09 - Sensitivity Calculation to Account for Uncertainty Background/Issue: EPRI TR 3002025288 section 4.3 states that analysis using a factor of 3 reduction in reliability for systems categorized as RISC-3 is conservative and appropriate, citing NEI 00-04. It is unclear why this factor is conservative and appropriate in the reversed context of this methodology, where components are presumed LSS by default, in contrast to the traditional 50.69 methodology which presumes components are HSS by default. Notably, the proposed methodology is relatively simplified compared to the traditional use of NEI 00-04 for supporting 10 CFR 50.69 applications which includes a relatively fine-grained assessment of subject systems.

NEI 00-04 does not state that a factor of 3 is appropriate, rather it provides a range of values useful in conducting sensitivity studies of an analysis. No basis is given for this range in NEI 00-04 beyond that it would provide trend insights for the consequences of reductions in reliability due to reduced treatments. The factor of 3 is generally used when assessing sensitivity to uncertainty as it is an approximation of the likely tail of a distribution for active systems. When altering the general approach (e.g. changing from HSS treatment to LSS treatment for passive systems), it is unclear why it is reasonable to assess the future distribution (LSS treatment) as matching the prior distribution (HSS treatment).

Addressing this uncertainty is particularly important in the context of other relaxations in treatment that may occur due to changes in ASME code requirements, for example, that may be implemented separately and concurrently with this methodology. This is particularly important in understanding whether the factor chosen genuinely informs regarding uncertainties in the context of passive systems and the performance monitoring associated with such.

The NRC staff needs a justification of the use of a factor of 3 provided on the docket.

Requests:

a.

Clarify on what basis a factor of 3 is determined to be conservative. In particular, provide any operating experience meta-analysis and/or data distributions supporting that a factor of 3 is conservative, or realistic for passive systems.