DCL-24-075, Response to Request for Additional Information for License Amendment Request 23-02, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power React

Response to Request for Additional Information for License Amendment Request 23-02, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power React

ML24221A140
Person / Time
Site:	Diablo Canyon   (DPR-080, DPR-082)
Issue date:	08/08/2024
From:	Rogers J Pacific Gas & Electric Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
DCL-24-075
Download: ML24221A140 (1)

Text

Justin E. Rogers Station Director Diablo Canyon Power Plant Mail code 104/5/502 P.O. Box 56 Avila Beach, CA 93424 805.545.3088 Justin.Rogers@pge.com A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek PG&E Letter DCL-24-075 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Diablo Canyon Units 1 and 2 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Response to Request for Additional Information for License Amendment Request 23-02, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Reference 1: PG&E Letter DCL-23-07, License Amendment Request 23-02 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated September 27, 2023 (ADAMS Accession No. ML23270B909) 2: NRC email Request for Additional Information by the Office of Nuclear Reactor Regulation Diablo Canyon 50.69 Risk-Informed Categorization, dated July 2, 2024 (ADAMS Accession No. ML24184C042)

Dear Commissioners and Staff:

Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) submitted Reference 1 that requested approval of a proposed amendment to modify the Diablo Canyon Power Plant (DCPP) Units 1 and 2 licensing bases, by the addition of a License Condition, to allow for the implementation of the provisions of 10 CFR, Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. In Reference 2, the staff provided a request for additional information to support the staff review. This letter provides the response to the requested additional information.

The response in the Enclosure does not impact the significant hazards evaluation or environment evaluation contained in Reference 1.

PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this letter.

m PacHic Gas and Electric Company*

Document Control Desk PG&E Letter DCL-24-075 Page 2 A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek Pursuant to 10 CFR 50.91, PG&E is sending a copy of this letter to the California Department of Public Health.

If you have any questions or require additional information, please contact James Morris, Regulatory Services Manager, at 805-545-4609.

I state under penalty of perjury that the foregoing is true and correct.

Sincerely, Justin E. Rogers Station Director Executed on: ____________________

Date kjse/ SAPN 51245309 Enclosure cc:

Diablo Distribution cc/enc: Anthony Chu, Branch Chief, California Dept of Public Health Mahdi O. Hayes, NRC Senior Resident Inspector Samson S. Lee, NRR Project Manager John D. Monninger, NRC Region IV Deputy Administrator 08/08/2024

Enclosure PG&E Letter DCL-24-075 1 of 8 Response to Request for Additional Information By the Office of Nuclear Reactor Regulation Diablo Canyon 50.69 Diablo Canyon 50.69 Risk-Informed Categorization

1.

Background

In a letter dated September 27, 2023, Pacific Gas and Electric Company (the licensee) requested an amendment to its license for Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Diablo Canyon). The licensees proposed amendment would add license conditions for each unit to allow Diablo Canyon to implement Title 10 of the Code of Federal Regulations (10 CFR) section 50.69, Risk-informed categorization and treatment of structures, systems and components [SSCs] for nuclear power reactors.

The provisions of 10 CFR 50.69 allow licensees to use an integrated, systematic, risk-informed process for categorizing SSCs according to their safety significance. A licensee that has adopted 10 CFR 50.69 may specify alternative treatments for SSCs that have low safety significance.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information in the license amendment request (LAR) and determined that additional information is required to complete its review. The NRC staffs requests for additional information (RAIs) are provided in section 3 below.

2.

Regulatory Basis Special treatment requirements are imposed on safety-related SSCs of a nuclear power plant to provide increased assurance (beyond normal industrial practices) that the SSCs are capable of meeting their functional requirements under design-basis conditions. These requirements go beyond the controls and measures typically applied to equipment classified as commercial grade. These additional requirements include design considerations, qualification, change control, documentation, reporting, maintenance, testing, surveillance, and other quality assurance requirements.

Licensees may voluntarily adopt 10 CFR 50.69 to implement an alternative regulatory framework with respect to requirements to provide adequate assurance that SSCs of low safety significance will perform their design-basis functions.

The Nuclear Energy Institute (NEI) issued guidance for implementation of a process for categorizing SSCs: NEI 0004, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (NEI 0004).

The NRC issued, for trial use, Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance (RG 1.201). It endorses NEI 0004 with clarifications and limitations.

RG 1.201 describes a method that the NRC staff considers acceptable for the categorization of SSCs that are considered in risk-informing special treatment requirements. Use of this method complies with the Commissions requirements in 10 CFR 50.69.

Enclosure PG&E Letter DCL-24-075 2 of 8

3.

RAIs

RAI 1

The LAR states that the models used in the probabilistic risk assessment (PRA) were peer reviewed using ASME/ANS RASa-2009 and RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities. For the seismic PRA, however, the LAR states that a full-scope seismic PRA peer review, which also included a review of the seismic hazard and fragility analyses, was conducted in June 2017, and it was performed consistent with this revision of RG 1.200, but using ASME/ANS RASb-2013. The LAR states that an independent assessment of the finding-level facts and observations (F&Os) was conducted from October to December 2017 and the scope of the assessment included all finding-level F&Os resulting from the peer review. The LAR also states that a focused-scope peer review was conducted in conjunction with the closure review and that there are no remaining open peer review finding level F&Os.

The NRC staff notes that RG 1.200, Revision 2, endorses ASME/ANS RASa2009, but it does not endorse ASME/ANS RASb2013. Similarly, the NRC staff notes that RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, does not endorse ASME/ANS RASb2013. As discussed in RG 1.200, Revision 2, a risk-informed submittal should contain discussions concerning peer review. If the peer review is not performed against the endorsed standards, RG 1.200, Revision 2, states that information needs to be included in the submittal that demonstrates that the different criteria used are consistent with the endorsed standards.

The NRC staff notes that this issue was discussed during the NRC staffs audit for the Diablo Canyon LAR to adopt risk-informed completion times dated July 13, 2023. The license provided a comparison of the criteria in ASME/ANS RASb2013 with the criteria in ASME/ANS RASa2009 in a letter dated January 15, 2024.

Please address the following:

1. Confirm that the comparison of the criteria in ASME/ANS RASb2013 with the criteria in ASME/ANS RASa2009 in the letter dated January 15, 2024, is valid for this LAR. If the comparison is not valid for this LAR, then provide a comparison of the criteria in ASME/ANS RASb2013, which has not been endorsed by the NRC for licensing applications, with the criteria in the endorsed ASME/ANS RASa2009, including an explanation that demonstrates that the analogous ASME/ANS RASa2009 supporting requirements have been met for instances where the criteria differ between the two standards.

PG&E Response:

The response to Audit Question APLC-02 (a) in PG&E letter DCL-24-004, Supplement to License Amendment Request 23-01 Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b', dated January 15, 2024, is valid for the Diablo Canyon 50.69 LAR.

Enclosure PG&E Letter DCL-24-075 3 of 8

RAI 2

Paragraph (c)(1)(ii) of 10 CFR 50.69 requires that the SSC functional importance be determined using an integrated, systematic process. NEI 00-04, Section 5.6, Integral Assessment, discusses the need for an integrated computation using available importance measures. It further states that the integrated importance measure essentially weighs the importance from each risk contributor (e.g., internal events, fire, seismic PRAs) by the fraction of the total core damage frequency [or large early release frequency] contributed by that contributor. The guidance provides formulas to compute the integrated Fussell-Vesely importance (FV) and integrated Risk Achievement Worth (RAW).

Based on the information provided in the LAR, it is not clear to the NRC staff how the licensee proposes to address the integration of importance measures across all hazards (i.e., internal events, internal flooding, fire, and seismic). Please address the following:

1. Explain how the integration of importance measures across hazards for the 10 CFR 50.69 categorization process will be performed.

PG&E Response:

As noted in Section 3.1.1 of PG&E letter to NRC, DCL-23-077, Diablo Canyon, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, dated September 27, 2023, The process to categorize each system will be consistent with the guidance in NEI 00-04, '10 CFR 50.69 SSC Categorization Guideline,' as endorsed by RG 1.201. The approach described in Section 1.5 of NEI 00-04 regarding Integrated Importance Assessment states, In order to facilitate an overall assessment of the risk significance of SSCs, an integrated computation is performed using the available importance measures. This integrated importance measure essentially creates a weighted-average importance based on the importance measures and the risk contributed by each hazard (e.g., internal events, fire, seismic PRAs). A weighted average using the importance measures and corresponding core damage frequency (CDF) and Large Early Release Frequency (LERF) from each of the hazards will be used as described in NEI 00-04.

2. Discuss how the importance measures for the PRA models (e.g., FV and RAW) are derived and justify why the importance measures generated do not deviate from the NEI guidance or Table 3-1 of the LAR. If the practice or method used to generate the integrated importance measures is determined to deviate from the NEI guidance, then provide justification to support why the integrated importance measures computed are appropriate for use in the categorization process.

PG&E Response:

The process used to derive the importance measures for various hazards as described in Table 3-1 of letter DCL-23-077 is performed on a component basis in accordance with guidance in NEI 00-04 Section 5 "Component Safety Significance Assessment". Some components in the Internal Events PRA may not be modeled explicitly in the Seismic PRA due to it being

Enclosure PG&E Letter DCL-24-075 4 of 8 subsumed in a super component (as described in NEI 00-04), or because it was screened out of the analysis. For those that were screened out, the importance measure would be taken as non-risk significant (0 for FV or 1 for RAW). For the components that are included under a super component, the super component risk importance would instead be considered to determine risk significance. This treatment for components included under a super component is consistent with the NEI 00-04 Section 5 sentence "As outlined in Section 1, by focusing on the significance of system functions and then correlating those functions to specific components that support the function, it is possible to address even implicitly modeled components."

3. Describe how the importance measures for the seismic PRA (e.g., FV and RAW) are derived considering that the seismic hazard is discretized into bins. The discussion should include how the same basic events, which were discretized by binning during the development of the seismic PRA, are then combined (i.e., combined across bins as well as across failure modes such as seismic and random failure modes) to develop representative importance measures. Further, discuss how they are compared to the importance measure thresholds in NEI 00-04. Provide justification to support the determined impact on the categorization results and describe how the approach is consistent with the guidance in NEI 00-04.

PG&E Response:

PG&E will take each of the importance measures and create a weighted average based on its CDF/LERF weighting as described in NEI 00-04 Section 1.5 Integrated Importance Assessment and Section 5.6 Integral Assessment. Therefore, derivation of the importance measures is consistent with NEI 00-04 guidance.

4. In the context of the integral assessment described in NEI 00-04, Section 5.6, it is understood that importance evaluations performed in accordance with the process in NEI 00-04 are determined on a component basis. However, the LAR and NEI 00-04 guidance does not make clear how the integrated importance measures are calculated for certain components. Specifically, in the seismic PRA, basic events that represent different failure modes for a component may not align with basic events in other PRA models.

Examples of such basic events include those that are specific to the seismic PRA (including implicitly modeled components) or basic events that represent a subcomponent modeled within the boundary of a component in the internal events PRA.

Provide details and justification to support how the integrated importance measures will be calculated for the basic events modeled in the seismic PRA that may not align directly with basic events modeled in the PRA for other hazards. Include discussion for any mapping that will be performed across the seismic PRA basic events and those in other PRA modeled hazards where additional modelling is determined to be necessary.

PG&E Response:

The process employed for Diablo Canyon is justified because the importance evaluations are performed in accordance with NEI 00-04, which are determined on a component basis. Some

Enclosure PG&E Letter DCL-24-075 5 of 8 components in the Internal Events PRA may not be modeled explicitly in the Seismic PRA due to it being subsumed in a super component (as described in NEI 00-04), or because it was screened out of the analysis. For those that were screened out, the importance measure would be taken as non-risk significant (0 for FV or 1 for RAW). For the components that are included under a super component, the super component risk importance would instead be considered to determine risk significance.

RAI 3

In 10 CFR 50.69(c)(1)(i) and (ii), the regulations require that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process and 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.

Industry guidance (NEI 0004) states that sensitivity studies should be conducted to address key assumptions. Sensitivity studies on human error rates, common-cause failures, and maintenance unavailabilities are performed to ensure that assumptions of the PRA are not masking the importance of an SSC. The guidance also recommends the use of sensitivity studies identified in the characterization of PRA adequacy if they apply.

The LAR states that assumptions and sources of uncertainty were reviewed to identify those that would be significant in the risk-informed categorization process. In attachment 5 to the enclosure, the licensee provided a table that summarized key assumptions and sources of uncertainty with a discussion of how each one was or will be addressed. In several cases, additional information is needed for the staff to confirm that the documented dispositions will satisfy the requirements of 10 CFR 50.69.

1. The LAR states that dual unit trips (except for seismic events) are not considered in the single unit model, and crosstie to the other unit's resources may be unavailable. Moreover, it states that sensitivity studies will be performed for the affected SSCs.

Describe how the interdependence of structures and systems of the opposite unit will be addressed when conducting sensitivity analysis of shared components. Justify the adequacy of this approach for the categorization results.

PG&E Response:

Shared systems as described in Section 1.2.2.10 of the Updated Final Safety Analysis Report modeled in the PRA include Auxiliary Saltwater (ASW) pumps, electrical support for ASW (Diesel Generators, 4.16 Kilovolt buses, 125 Volt direct current power), and Diesel Fuel Oil.

Implementation of 50.69 will address the interdependence of structures and systems of the opposite unit by including cross unit impacts in the model that will be used for categorization results. For those structures and systems that impact both units, implementation of 50.69 will assess the impact of both units for categorization of the respective units. However, crosstie components from the other unit will be disabled when performing sensitivity analyses.

Enclosure PG&E Letter DCL-24-075 6 of 8

2. The LAR states that for Charging and Safety Injection pumps credited in a medium loss-of-coolant accident, it was assumed that 2 out of 4 high-pressure pumps are required for success. It further states that this was conservatively modeled as 1 out of 2 Charging pumps and 1 out of 2 Safety Injection pumps.

a. Confirm that any two of the four high-pressure pumps are sufficient or explain the actual success criteria in more detail.

PG&E Response:

Charging and Safety Injection pumps are credited for inventory make-up in case of a medium loss-of-coolant accident (LOCA). DCPP PRA success criteria requires that any 2 out of the 4 Safety Injection or Charging pumps function (Charging or Safety Injection) to mitigate a medium LOCA; this was conservatively modeled as 1 of 2 Charging pumps and 1 of 2 Safety Injection pumps. This conservative modeling approach is required due to how these systems are modeled in the DCPP Riskman medium LOCA event tree which includes separate top events for Safety Injection and Charging. With Charging and Safety Injection pumps modeled in separate top events, the model cant distinguish between a success state where two pumps are available versus a success state where only one pump is available.

b. Explain why 2 of 2 charging pumps might be required and justify the method used to preserve conservatism in this case.

PG&E Response:

The medium LOCA recovery model was used to recover specific medium LOCA scenarios. It was assumed that 2 of 4 high-head pumps were required for success; this was conservatively modeled as 1 of 2 Charging pumps and 1 of 2 Safety Injection pumps. If the Charging system is found to be successful (at least 1 of 2 Charging pumps) but both Safety Injection pumps failed, failure of high head injection was assumed. To address this conservatism, when the Charging top event is successful and the Safety Injection top event is failed, a recovery factor equal to 2 times the charging pump train failure probability is used to approximate the reliability of two Charging pumps.

c. Explain why this approach does not affect the categorization results.

PG&E Response:

This approach is unlikely to have an impact on the categorization results due to the insignificant contribution to CDF from cutsets involving this success criteria. Scenarios that utilize this success criteria involve a low likelihood initiator (Medium LOCA frequency is 2.3-05/yr) and a low likelihood of Safety Injection failure (approximately 7E-03 failure probability). When combined, the total cutset frequency is approximately 1.6E-07/yr. To confirm that categorization

Enclosure PG&E Letter DCL-24-075 7 of 8 is unaffected by this modeling assumption, an assessment will be performed as part of each system categorization.

3. The LAR states that reduction of the mission time from 24 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the diesel generators does not have a significant impact in the baseline PRA. Justify the conclusion that this has no significant impact on the categorization results.

PG&E Response:

A convolution method electric power recovery model is used to quantify the probability distribution for the cumulative onsite power failure probability at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the probability distribution for the onsite power failure and electric power non-recovery probability integrated over the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the loss of offsite power. As such, the full 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time is accounted for in the approach to modeling electric power recovery. This modeling approach does not impact the categorization results.

4. The LAR states an assumption that vacuum breakers cannot fail in a manner that has an adverse impact on Auxiliary Salt Water (ASW) function. It also states that the uncertainty attributable to this this nonconservative assumption is not known. Justify the expectation that the contribution of this assumption is small and does not significantly affect the categorization results.

PG&E Response:

The ASW vacuum relief valves are not modeled in the PRA because they have an insignificant contribution to ASW system reliability and to overall risk (they have been demonstrated to have a failure probability two orders of magnitude below the highest failure probability of other components in the same train). These valves are highly reliable mechanical components and, ASW train failure would only occur if both vacuum relief valves were to fail.

Because of their low risk significance, the vacuum relief valves would not have an impact on system categorization.

5. The LAR states that certain SSCs are always failed in the fire PRA and seismic PRA models. Briefly identify which systems are handled in this manner and why treating them conservatively will not adversely affect SSC categorization.

PG&E Response:

Systems and components assumed failed in the Seismic and Fire PRA models include:

500-kV offsite power system Non-vital power systems Unit bus crosstie breaker Opposite unit startup power crosstie

Enclosure PG&E Letter DCL-24-075 8 of 8 Balance of plant systems including main feedwater, condensate, instrument air, circulating water, service cooling water Anticipated Transient Without Scram Mitigating System Actuation Circuitry System Containment Spray System Containment Fan Cooler Units Makeup from the Spent Fuel Pool System For each 50.69 system categorization, PG&E will assess the impact that this conservative assumption has on categorization. An example approach to perform this assessment would be to perform a sensitivity that removes the failed component impacts to understand the categorization impact of the modeling assumption. The results of this assessment would then be considered by the Independent Decision Making Panel as part of the categorization process.