ML25183A016

CFR 50.69 Audit Questions (Apla) Dated July 1, 2025

ML25183A016
Person / Time
Site:	Wolf Creek
Issue date:	07/01/2025
From:	Samson Lee Plant Licensing Branch IV
To:	Lee N Wolf Creek
Lee S, 301-415-3168
References
EPID L-2025-LLA-0017
Download: ML25183A016 (1)

Text

From:

Samson Lee To:

Nathan Lee

Subject:

Wolf Creek 10 CFR 50.69 audit questions (APLA) dated July 1, 2025 Date:

Tuesday, July 1, 2025 12:56:00 PM By letter dated January 30, 2025, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25030A384), Wolf Creek Nuclear Operating Corporation (the licensee) submitted a license amendment request (LAR) for the Wolf Creek Generating Station to adopt Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors.

The U.S. Nuclear Regulatory Commission (NRC) staff has identified the need for a regulatory audit to examine the licensees non-docketed information with the intent to gain understanding, to verify information, or to identify information that will require docketing to support the basis of the licensing or regulatory decision. The NRC staff issued an audit plan dated April 2, 2025 (ML25090A217). The NRC staff has formulated initial audit discussion questions below for certain technical areas. If time allows, please prepare responses to these questions in advance. It would facilitate the audit discussions, especially if responses can be posted in the online portal as they become available. The NRC staff plans to issue additional audit discussion questions for other technical areas to support breakout audit meetings for efficiency.

Wolf Creek 50.69 Audit Questions (APLA, July 1, 2025)

PRA [Probabilistic Risk Assessment] Licensing Branch A (APLA)

APLA-01: Interfacing System Categorization

Section 7.1 of Nuclear Energy Institute (NEI) 00-04 states, "[d]ue to the overlap of functions and components, a significant number of components support multiple functions. In this case, the system, structure, and component (SSC), or part thereof, should be assigned the highest risk significance for any function that the SSC or part thereof supports." Section 4 of NEI 00-04 states that a candidate low safety-significant (LSS) SSC that supports an interfacing system should remain uncategorized until all interfacing systems are categorized. The license amendment request (LAR) does not discuss consideration or implementation of the guidance in Section 7.1 of NEI 00-04.

Explain how the categorization process will be implemented to ensure that the cited guidance in NEI 00-04 will be followed and that any functions/SSCs that serve as an interface between two or more systems will not be categorized until the categorization for all of the systems that they support is completed and that SSCs that support multiple functions will be assigned the highest risk significance for any of the functions they support.

APLA-02: Determination of Key Sources of Uncertainty for the 10CFR50.69 Categorization Process and Sensitivity Results

Sections 50.69(c)(1)(i) and 50.69(c)(1)(ii) of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.69 require that a licensees probabilistic risk assessment (PRA) be of sufficient quality and level of detail to support the SSC categorization process, and that 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. The guidance in NEI 00-04 specifies that sensitivity studies be conducted for each PRA model to address uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask the importance of components. The guidance in NEI 00-04 states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.

of the LAR describes the process used for reviewing the PRA assumptions and sources of uncertainty. The submittal states: The Internal Events and Fire PRAs identify assumptions and determine if those assumption are related to sources of model uncertainty and characterize that uncertainty, as necessary. The NRC staff reviewed the uncertainty documents provided on this audits electronic portal for the internal events, internal flooding, and fire PRA and found that further clarification is necessary regarding the review of assumptions and sources of uncertainty for this application. It is unclear if additional analysis was performed and documented to determine if any source of uncertainty could adversely impact any SSC categorization. In light of these observations, provide the following information:

a. Provide details of how the Wolf Creek PRA sources of uncertainty were evaluated as a potential key source of uncertainty for this application. In this response provide any documentation of this process.

b. Provide the results of sensitivity studies that determined the impact on risk for each associated source of uncertainty. Include in this discussion justification that the sensitivity results demonstrate that the associated source of uncertainty does not adversely impact any SSC categorization.

Docket No. 50-482