CP-202300575, (Cpnpp), License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Supplement 2

(Cpnpp), License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Supplement 2

ML23347A206
Person / Time
Site:	Comanche Peak
Issue date:	12/13/2023
From:	John Lloyd Luminant, Vistra Operating Co. (VistraOpCo)
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
CP-202300575, TXX-23094
Download: ML23347A206 (1)

Text

Jay J. Lloyd Comanche Peak Senior Director, Nuclear Power Plant Engineering & Regulatory Affairs (Vistra Operations Company LLC)

P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043

T 254.897.5337

CP-202300575 TXX-23094 December 13, 2023

U. S. Nuclear Regulatory Commission Ref 10 CFR 50.90 ATTN:DocumentControlDesk Washington, DC 20555-0001

Subject:

Comanche Peak Nuclear Pow er Plant (CPNPP)

Docket Nos. 50-445 and 50-446 License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment Strof uctures Sy, stems an Cod mponents for Nuclear Power Reactors Supplement 2

References:

1. Letter, J. Lloyd to NRC, Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, sy stems and components for nuclear power reactors. LAR 23-001, April 19, 2023, ML23109A333

2. Letter, J. Lloyd to NRC, Response to Request for Supplemental Information for License Amendment Request to Adopt 10 CFR 50.59, R isk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2023-LLA-0057), June 8, 2023, ML23159A200

Dear Sir or Madam:

By References 1 and 2, Vistra Operations Company LLC (Vistra OpCo) submitted a license amendment request (LAR) to adopt 10 CFR 50.69, Risk-inform ed categorization and treatment of structures, systems and components for nuclear power reactors. NRC staff conducted a regulatory audit to support review of the proposed LAR. Vistra OpCo provides in Enclosure 1 of this letter a supplement to the proposed LAR in response to the regulatory audi t. Enclosure 2 of this letter provides a revised license condition that replaces the license conditio ns previously proposed in References 1 and 2.

This communication contains no new commitments regarding CPNPP Units 1 and 2.

Should you have any questions, please contact N. Boehmisch at (254) 897-5064 or nicholas.boehmisch@luminant.com TXX-23094 Page 2 of 2

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

Executed on December 13, 2023.

Sincerely,

Jay J. Lloyd

Response to Audit Questions : Revised License Condition

c (email) - John Monninger, Region IV [John.Monninger@nrc.gov]

Dennis Galvin, NRR [Dennis.Galvin@nrc.gov]

John Ellegood, Senior Resident Inspector, CPNPP [John.Ellegood@nrc.gov]

Dominic Antonangeli, Resident Inspector, CPNPP [Dominic.Antonangeli@nrc.gov]

to TXX-23094 Page 1 of 33

AUDIT QUESTIONS LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED

CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS

FOR NUCLEAR POWER REACTORS

VISTRA OPERATIONS COMPANY, LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2

DOCKET NO. 50-445 AND 50-446

By letter dated April 19, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23109A333), as supplemented by letter dated June 8, 2023 (ML23159A200), Vistra Operations Company LLC (Vistra OpCo, the licensee) submitted a license amendment request (LAR) for the use of a risk-informed process for the categorization and treatment of structures, systems, and components at Comanche Peak Nuclear Power Plant (Comanche Peak or CPNPP), Units 1 and 2. The proposed LAR would modify the Comanche Peak licensing basis, by the addition of a License Condition, to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." The U.S. Nuclear Regulatory Commission (NRC) staff have reviewed the LAR and information posted on the online portal and have identified the following audit questions.

APLA Question 01 - Open Internal Event Findings and Observations (F&Os)

Sections 50.69(b)(2)(ii) of 10 CFR 50.69 requires the PRA [probabilistic risk assessment] 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. Regulatory Guide (RG) 1.200, Revision 3 An Approach for Determining the Technical Adequacy of Probabili stic Risk Assessment Results for Risk-informed Activities, (ML20238B871), provides guidance for addressing PRA acceptability. RG 1.200 describes a peer review process utilizi ng the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA standard (currently ASME/ANS-RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established for evaluations that could influence the regulatory decision. RG 1.200, Revision 3 states that, in general, it is expected that the majority off applications meet Capability Category (CC) is II, but CC-I may be acceptable for some requirements. A process to close-out Finding-level F&Os is documented in NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard (ML19231A182) that is endorsed by RG 1.200, Revision 3.

In the supplement to the LAR dated June 8, 2023, the licensee acknowledged there are two supporting requirements (SRs) (i.e., IFEV-A6 and LE-C11)) that are met only at CC-I and provided assessments of their impact on the 10 CFR 50.69 application. Concerning LE-C11, the supplement states that [n]o credit was taken for continued operation after containment failure and that not crediting such actions has minimal impact on 10 CFR 50.69 categorization because in these cases the core is damaged, and containment has failed. NRC staff observes that crediting operator actions and/or other equipment after containment failure could potentially limit the release to the environment from the core to less than a Large Early Release for certain to TXX-23094 Page 2 of 33

scenarios and meaningfully skew the large early release frequency (LERF) importance factors.

Given the conclusion that the CC-I modeling associated with LE-C11 is conservative and has no impact on 10 CFR 50.69 categorization, the implication seems to be that no further evaluation is needed. However, the supplement also appears to indicate that the impact associated with CC-I modeling would be evaluated under the 10 CFR 50.69 program at the time of categorization (e.g., the statement impact on specific applications will be evaluated as needed seems to imply this). Given these observations and potential inconsistencies provide the following:

a) Clarify whether not crediting continued operation of mitigating equipment and operator actions after containment failure will be evaluated at the time of structure, system, and component (SSC) categorization. If so, explain how this will be performed (e.g., using a sensitivity study) and what mechanism will be used to ensure that this CC-I modeling is evaluated.

b) If the CC-I modeling associated with SR LE-C11 described above is not to be further evaluated, then justify it has minimal impact on CFR 50.69 categorization. Include discussion of a basis for concluding that the conservative CC-1 associated with SR LE-C11 would not meaningfully skew importance measure values determined used in the 10 CFR 50.69 categorization process for LERF.

Response

It is understood that not crediting continued operation of mitigating equipment and operator actions after containment failure can skew the importance of containment related equipment towards LERF. Crediting post core melt actions may result in some releases falling into late categories or even the intact category.

There is limited equipment credited in the CPNPP PRA model associated with the Level II assessment - primarily the containment sp ray and containment isolation components and supports. The CPNPP Level 2 results are dominated by by-pass and steam generator tube rupture sequences. Containment over-pressurization scenarios were found to be low contributors due to the large dry containment size of CPNPP as evidenced by the low significance of the containment spray system. SGTR scenarios and releases are influenced by the availability of the Auxiliary Feedwater System, which is identified as safety significant for CDF or, for LERF, would bound its importance if actions were credited to shift the release bins.

With regards to containment isolation components, the size of the penetration plays a more significant role in determining the appropriate Level 2 bin. This binning (by large vs small release, and the late vs early release) and their contribution to any of the release bins will be considered in the supporting categorization assessment.

Therefore, the CC-I modeling associated with SR LE-C11 will not play a major role in identifying significant Level 2 components.

The 50.69 Program implementing procedures will incl ude instructions to the categorization team to include an assessment of these effects in the information package for consideration of the Integrated Decision-making Panel.

to TXX-23094 Page 3 of 33

APLA Question 02 - Determination of Key Sources of Uncertainty for the 10 CFR 50.69 Categorization Process

Sections 50.69(c)(1)(i) and 50.69(c)(1)(ii) of 10 CFR 50.69 require that a licensees 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 10 CFR 50.69 SSC Categorization Guideline, 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 certain components. The guidance in NEI 00-04 states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.

The supplement to the LAR dated June 8, 2023, presents Table 1: Disposition of Key Sources of Assumptions and Uncertainties Related to 50.69. The evaluations in Table 1 indicates that the treatment of the identified sources of uncertainty is either realistic or somewhat conservative. The supplement states that the conclusion of the review is that no additional sensitivity analyses are required to address the CPNPP PRA model specific assumption or sources of uncertainty for [10 CFR] 50.69. Given that key sources of uncertainty should be evaluated using sensitivity studies per the guidance in NEI 00-04, it is inferred that the licensee is concluding that there are no key sources of uncertainty associated with the PRA for this application. Accordingly confirm that there are no key sources of uncertainty associated with the PRA for the 10 CFR 50.69 application.

Response

The response provided in the cited LAR supplement remains valid.

to TXX-23094 Page 4 of 33

APLA Question 03 - Open Phase Condition (OPC)

Section C.1.4 of RG 1.200 states the base (e.g., Model of Record (MOR)) PRA is to represent the as-built, as-operated plant to the extent needed to support the application. The licensee is to have a process that identifies updated plant information that necessitate changes to the base PRA model.

The staff noted that in the Comanche Peak TSTF-505 LAR and its supplements (see references 1 to 6 in ML22192A007), the licen see indicated that the PRA model used to support TSTF-505 application did not credit the Open Phase Isol ation System (OPIS) system or associated operator actions. Furthermore, the potential for an OPC event and OPIS modification evaluation are identified for reference in model maintenance documentation. It is not clear to the staff whether the OPIS system or associated operator actions are not credited for the 10 CFR 50.69 application.

Clarify whether the OPIS system or associated operator actions are credited for the 10 CFR 50.69 application. If they are credited for the 10 CFR 50.69 application, address the following:

a) Discuss the impact on risk of the OPC issue at Comanche Peak, Units 1 and 2.

b) Discuss whether modeling of the OPC issue and any OPIS that has been installed and implemented at Comanche Peak have been, or are planned to be, incorporated as part of the plant MOR. If so, provide the following:

i. The schedule for the inclusion of OPC and OPIS modeling in the MOR.

ii. The impact, if any, to key assumptions and sources of uncertainty.

iii. A discussion of the human reliability assessment (HRA) methods and assumptions used for OPIS alarm manual response.

iv. The impact to external events (e.g., fire, seismic, flooding, high winds, tornado, other external events, etc.).

v. A discussion of the risk impact of inadvertent OPIS actuation and justification for its exclusion.

c) If OPC and OPIS are not planned to be included in the MOR, provide just ification why the risk impact is not included by performing either a qualitative or sensitivity analysis.

Response

The OPIS is not explicitly modeled in the Coma nche Peak PRA models.

a) Evaluation EV-TR-2019-006419-6 (Evaluation Log #310) includes a general descripti on of the OPEN PHASE PROTECTION (OPP) SYSTEM installed on each of CPNP P's four Startup Transformers, details the e xpected plant response to an OPC and documents the plant specific risk evaluation. A plant modification installed a dual channe l OPP system at each transformer to monitor for the condition and prompt an operator response. The function of the OPP, as described in the system specification, is to detect the condition,

actuate protective features, and alarm remotely when open circuit faults on offsite po wer to TXX-23094 Page 5 of 33

sources occur under unloaded and loaded transformer operating conditions. For CPNPP, an OPC does not immediately cause a plant initiating event. OPC frequency and probability estimates were obtained from the available published estimates (NEI 19-02).

These values were judged to be applicable; using a plant capacity factor of 0.92, an OPC Frequency= 7.5 events in 15 calendar years = 7.5 I (0.92 *1500 reactor-years) = 5.43E-03 per reactor-year was applied as representative for the purpose of evaluating the OPC impacts associated with the CPNPP plant response. Because the system is implemented as Alarm Only, OPC initiating plant trips have the same SSC consequences as an interruption of AC power with OPC effects limited to the direct bus the transformer serves.

Operator actions are established to detect or determine an OPC condition exists and restore 6.9kV safeguards bus balanced voltage, by first, opening supply breakers and second, ensuring transfer of at least one safeguards bus. An operator recovery action provides for manual start of equipment that is running or demanded, if the equipment were to trip when running or demanded during an OPC.

b) The PRA model used to support this application did not model the OPIS system or associated operator actions. The potential for an OPC event and OPIS modification evaluation are identified for reference in model maintenance documentation.

c) Exclusion of this failure mode and mitigat ing system from the PRA Model does not impact the risk significance calculations. The plant is assumed to be in the normal electrical configuration with more than one transmission feeder aligned to the switchyard. Time spent in unusual configurations which would propagate a phase imbalance via an OPC in the transmission system is assumed to be small. An OPC event is similar in nature to plant/grid centered LOOP modeled in the internal events PRA; the loss of power from an OPC condition is bounded by these initiators for frequency and impact to station startup transformers. Safeguard Class 1 E AC 6.9kV buses (1 EA 1, 1 EA2) power frontline safety systems (required after a transient or initiating event) and as they are mitigating systems in standby mode, a loss of their power source will not result in a plant transient. The loss of a 1 E AC 6.9kV bus will lead to an orderly plant shutdown as required by technical specifications and does not meet the criteria to be considered an initiating event.

d) The OPC Incremental Core Damage Probability (ICDP) was developed by using fault tree logic corresponding to the NEI 19-02 event trees and reviewing the PRA modeling assumptions, application of OPC and OPIS failures, and the CPNPP electrical configuration. A review of other plant impacts and recovery as well as Human Failure Events associated with the response to an OPC was conducted. The OPC frequency and probability estimates were obtained from the available published estimates from NEI 19-02 and included the plant availability factor. This value, 5.43E-03 per reactor-year, was judged to be appropriate for application in the CPNPP PRA model. A representative action that models the operator actions required to separate the OPC from the emergency buses, given actuation of the OPIS alarm and recovery of equipment that is running or demanded if the equipment would trip when running or demanded during an OPC was then developed. Following a series of operator interviews, a Human Error Probability (HEP) value of 1.71E-03 was applied in the OPC analysis.

This information was then used to calculate the ICDP and ILERP. The table below provides results for the change in CDF and change in LERF risk metrics comparing the risk of operating the OPIS with the automatic trip function enabled versus operating with the alarm function only. to TXX-23094 Page 6 of 33

The OPC contribution to CDF is well below the maximum CDF contribution of <1E-6/yr.

This limiting value has a maximum ICDP impact that is well below the level of risk significance; therefore, the OPC impact would not impact any risk insights.

Exclusion of the OPIS system or associated op erator actions from the PRA model is not a key source of uncertainty for CPNPP. However, to ensure that the conclusion of the Open Phase Condition assessment remain valid, this modeling choice was added to the CPNPP initiating events and electric power notebooks such that this choice will be reviewed and assessed as the model is revised.

to TXX-23094 Page 7 of 33

APLA Question 04 - Periodic Review

10 CFR 50.69(e)(1) stated that in parts that The licensee shall review changes to the plant, operational practices, applicable plant and industry operational experience, and, as appropriate, update the PRA and SSC categorization and treatment processes. The licensee shall perform this review in a timely manner but no longer than once every two refueling outages.

In the LAR dated April 19, 2023, the licensee stated in part that Scheduled periodic reviews at least once every 48 months, consistent with the approved TSTF-505, Revision 2 identified in Section 3.2.4.1.7 in Reference 1, will evaluate new insights resulting from available risk information The staff noted that once every 48 months would exceed the rule requirement of no longer than once every two refueling outages.

Provide justifications to deviate from 10 CFR 50.69(e)(1) periodic review requirement.

Response

In accordance with 10 CFR 50.69(e)(1), CPNPP will review changes to the plant, operational practices, applicable plant and industry operational experience, and, as appropriate, update the PRA and SSC categorization and treatment processes. CPNPP will perform this review in a timely manner but no longer than once every two refueling outages.

Section 3.5 of the LAR is revised as follows to reflect the periodic review requirement.

3.5 FEEDBACK AND ADJUSTMENT PROCESS

If significant changes to the plant risk profile are identified, or if it is identified that a RISC-3 or RISC-4 SSC can (or actually did) prevent a safety significant function from being satisfied, an immediate evaluation and review will be performed prior to the normally scheduled periodic review. Otherwise, the assessment of potential equipment performance changes and new technical information will be performed during the normally scheduled periodic review cycle.

Scheduled periodic reviews, at least once every two refueling outages, will evaluate new insights resulting from available risk information (i.e., PRA model or other analysis used in the categorization) changes, design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated. This review will include:

• A review of plant modifications since the last review that could impact the SSC categorization

• A review of plant specific operating experience that could impact the SSC categorization

• A review of the impact of the updated risk information on the categorization process results

In addition to the normally scheduled periodic reviews, if a PRA model or the safe shutdown equipment lists are updated for any reason, including the 48-month updates required by the TSTF-505, R2 application, a review of the SSC categorization will be performed, including:

• A review of the importance measures used for screening in the categorization process

• An update of the risk sensitivity study performed for the categorization

The results of these assessments will be reviewed with the Integrated Decision-making Panel. to TXX-23094 Page 8 of 33

APLA Question 05 - Credit for FLEX Equipment and Actions

NRC memorandum dated May 6, 20221 provides the NRCs staff updated assessment of identified challenges and strategies for incorporating Diverse and Flexible Mitigation Capability (FLEX) equipment into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.2002. The staff noted that in the TSFT-505 LAR, the licensee indicated that FLEX equipment and associated mitigating actions are not credited at Comanche Peak. It is not clear to the staff whether FLEX equipment and associated mitigating actions are not credited for the 50.69 application.

Clarify whether FLEX equipment and associated mitigating actions are credited for the 50.69 application. If FLEX is credited for the 50.69 application, address the following:

a) Describe the methodology used to assess the failure probabilities of any modeled equipment credited in the licensee's mitigating strategies (i.e., FLEX). The discussion should include a justification of the rationale for parameter values, and how the uncertainties associated with the parameter values are considered in 50.69 categorization in accordance with ASME/ANS RA Sa-2009, as endorsed by RG 1.200 (e.g., supporting requirements for HLR-DA-D). This justification should also disposition any significant differences between these FLEX parameter values and those generic failure probabilities in PWROG-18042.

Alternatively, propose a mechanism to incorporate into the Comanche Peak PRA models updated FLEX parameter values prior to implementing 50.69.

b) Provide a discussion detailing the methodology used to assess operator actions related to FLEX equipment and the licensee personnel that perform these actions. The discussion should include:

i. A summary of the evaluation of the impact of the NRC clarifications in memorandum dated May 6, 2022, with regards to using the Electric Power Research Institute (EPRI) 3002013018 FLEX HRA methodology.

ii. Provide updated FLEX HRA results, if required, to address the NRC clarifications.

iii. Provide justification that the use of the EPRI FLEX HRA methodology does not meet the definition of an PRA upgrade as defined by RG 1.200.

Alternatively, if a justification is not provided, propose a mechanism to conduct a focused-scope peer review (FSPR) regarding incorporation of the EPRI FLEX HRA method for the Comanche Peak PRA models. Include in the mechanism the close-out all F&Os that result from the FSPR prior to implementing 50.69.

1 U.S. NRC memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, dated May 6, 2022 (ML22014A084).

2 U.S. Nuclear Regulatory Commission, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, RG 1.200, Revision 3, December 2020 (ML20238B871).

c) Provide an updated assessment of the impact on 50.69 SSC categorization by FLEX to TXX-23094 Page 9 of 33

equipment credited in Comanche Peak PRA models. This assessment should include, if required, any modifications to FLEX modeling based on the issues raised in this question. Include in this discussion, the impact of FLEX on exceeding importance rankings for affected SSCs and whether the uncertainty associated with FLEX modeling is a key source of uncertainty for 50.69. If this uncertainty is key, then describe and provide a basis for how this uncertainty will be addressed for 50.69 categorization.

Response

The CPNPP PRA models do not model any FLEX (portable) equipment or mitigating actions.

to TXX-23094 Page 10 of 33

APLC Question 01 - Alternate Seismic Approach

Paragraph (b)(2)(ii) of 10 CFR 50.69 requires, for license amendment, a description of measures taken to assure the level of detail of the systematic processes that evaluate the plant.

This includes the internal events at power PRA required by §50.69(c)(1)(i), as well as the risk analyses used to address external events.

In its supplement dated June 8, 2023, the licensee stated it will use EPRI Report 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, which is available in AD AMS (ML21082A170). The licensee stated that EPRI Report 3002017583 includes the updates to the Tier 1 approach resulting from the NRC staff's review of the Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (Calvert Cliffs), 10 CFR 50.69 amendment (ML19330D909). However, the staff finds that citing EPRI 3002017583 is not sufficient for Tier 1 seismic alternative approach, which also requires the licensee to provide information comparable to that provided by Calvert Cliffs.

The staff has previously requested and reviewed information to support its decision on the technical acceptability of the PRAs used in case studies as well as details of the conduct of the case studies. This information is included in the supplements to the Calvert Cliffs, LAR for adoption of 10 CFR 50.69. The supplement to the Calvert Cliffs 10 CFR 50.69 LAR dated May 10, 2019 (ML19130A180), contained additional information related to the alternate seismic approach including incorporation by reference of docketed information related to case study Plants A, C, and D; the supplement dated July 1, 2019 (ML19183A012), further clarified the information related to the alternate seismic approach (see response to Request for Additional Information (RAI) 4); the supplement dated July 19, 2019 (ML19200A216), provided responses to support the technical acceptability of the PRAs used for the Plant A, C, and D case studies as well as technical adequacy of certain details of the conduct of the case studies; the supplement dated August 15, 2019 (ML19217A143), clarified a response in the July 19, 2019, supplement.

The supplement dated July 19, 2019, included modifications to the content of the EPRI report.

In addition, the licensee removed several paragraphs related to its previous seismic submittals, categorization team evaluations, and integrated decision-making panels (IDP's) decision process from a typical Section 3.2.3.

Since the above-mentioned information was requested and reviewed by the staff for the Calvert Cliffs 10 CFR 50.69 LAR, the staff is unable to use it for the licensees docket unless it is incorporated in the licensees LAR. The above-m entioned information is necessary for the staff to make its regulatory finding on the licensees proposed alternate seismic approach and has not been provided by the licensee.

a) Provide the above-mentioned information to support the staffs regulatory finding on the alternate seismic approach by either incorporating the information by reference the identified supplements or responding to the RAIs in the identified supplements.

Response

To allow for reference in the future, sections shown at the bottom of this response are proposed to be modified compared to the original Comanche Peak Nuclear Generating Station, Units 1 and 2 (CPNPP) LAR (Reference 36) to allow for consistency with the Calvert Cliffs approach. Any modifications to the text shown also includes the update in the CPNPP supplement dated June 8, 2023 (Reference 37) for Response 2 which changed the EPRI methodology proposed for CPNPP from EPRI Report 3002022453 to ERPI Report 3002017583 (Reference 21). The updated sections incorporate updates, as appropriate, from the Calvert Cliffs supplement dated May 10, to TXX-23094 Page 11 of 33

2019 (Reference 38) and the Calvert Cliffs supplement dated July 1, 2019 (Reference 39).

Additionally, the Calvert Cliffs supplements discussed and responded to RAIs with regards to case studies identified in EPRI 3002012988 as shown in the supplement dated July 19, 2019 (Reference 40) and the supplement dated August 5, 2019 (Reference 41). These case studies and responses in the supplement are still valid for ERPI Report 3002017583 which is an updated ERPI 3002012988 to address RAIs at Calvert Cliffs as identified in the associated Calvert Cliffs supplements.

To that end and to support the proposed alternative seismic approach for CPNPP, Vistra OpCo incorporates by reference the following relevant information, with the differences identified in the response to APLC 01.b, that was provided to the NRC staff in docketed Calvert Cliffs Nuclear Power Plant 10 CFR 50.69 LAR correspondence. Incorporating relevant Calvert Cliffs Nuclear Power Plant 10 CFR 50.69 docketed correspondence by reference into this Vistra OpCo application for CPNPP is identical to the approach taken in the Hope Creek Generating Station 10 CFR 50.69 precedent for adopting the alternate seismic approach (Reference 35; see response to APLC RAI 05).

Calvert Cliffs Nuclear Power Plant LAR supplement dated May 10, 2019 (Reference 38);

contains additional information related to the alternate seismic approach including docketed information related to case study Plants A, C, and D.

Calvert Cliffs Nuclear Power Plant RAI response dated July 1, 2019 (Reference 39); further clarifies the information related to the alternate seismic approach (see Calvert Cliffs Nuclear Power Plant response to RAI 4).

Calvert Cliffs Nuclear Power Plant RAI response dated July 19, 2019 (Reference 40);

provides responses to support the technical acceptability of the PRAs used for the Plant A, C, and D case studies, as well as the technical adequacy of certain details of the conduct of the case studies (see Calvert Cliffs Nuclear Power Plant responses to RAI questions 1, 2, and 3).

Calvert Cliffs Nuclear Power Plant RAI response dated August 5, 2019 (Reference 41);

clarifies the response to RAI 3.

Note that the above Calvert Cliffs Nuclear Power Plant supplements incorporated by reference into this CPNPP application refer to EPRI Report 3002012988 instead of EPRI Report 3002017583. EPRI Report 3002017583, which is referred to in this LAR by CPNPP via the supplement (Reference 37), is a technical update that incorporated updates submitted to the NRC staff in a Calvert Cliffs 10 CFR 50.69 RAI response into EPRI Report 3002012988. Aside from those updates, the technical criteria in EPRI Report 3002017583 is identical to EPRI Report 3002012988

The following paragraphs in CPNPP LAR (Reference 36), Section 3.1.1 - Overall Categorization Process are shown below with the following modifications:

Original (in Reference 36) Modification The process to categorize each system will be The process to categorize each system will be consistent with the guidance in NEI 00-04, as consistent with the guidance in NEI 00-04, as endorsed by RG 1.201, with the exception of endorsed by RG 1.201, with the exception of the evaluation of impact of the seismic hazard, the evaluation of impact of the seismic hazard, which will use the EPRI Alternative Tier 1 which will use the EPRI Alternative Tier 1 Seismic Approach described in EPRI Report Seismic Approach described in EPRI Report 3002022453 (Reference 5). RG 1.201 states 3002017583 approach for seismic Tier 1 sites, that, the implementation of all processes which includes CPNPP, to assess seismic described in NEI 00-04 (i.e., Sections 2 hazard risk for 50.69. Inclusion of additional to TXX-23094 Page 12 of 33

through 12) is integral to providing reasonable process steps discussed below to address confidence and that all aspects of NEI 00-04 seismic considerations will ensure that must be followed to achieve reasonable reasonable confidence in the evaluations confidence in the evaluations required by required by 10 CFR 50.69(c)(1)(iv) is

§50.69(c)(1)(iv). However, neither RG 1.201 achieved. RG 1.201 states that, the nor NEI 00-04 prescribe a particular sequence implementation of all processes described in or order for each of the elements to be NEI 00-04 (i.e., Sections 2 through 12) is completed. Therefore, the order in which each integral to providing reasonable confidence of the elements of the categorization process and that all aspects of NEI 00-04 must be (listed below) is completed is flexible and as followed to achieve reasonable confidence in long as they are all completed, they may even the evaluations required by §50.69(c)(1)(iv).

be performed in parallel. Note that NEI 00-04 However, neither RG 1.201 nor NEI 00-04 only requires Item 3 to be completed for prescribe a particular sequence or order for components/functions categorized as LSS by each of the elements to be completed.

all other elements. Similarly, NEI 00-04 only Therefore, the order in which each of the requires Item 4 to be completed for safety elements of the categorization process (listed related active components/functions below) is completed is flexible and as long as categorized as LSS by all other elements: they are all completed, they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as LSS by all other elements. Similarly, NEI 00-04 only requires Item 4 to be completed for safety related active components/functions categorized as LSS by all other elements:

The risk analysis to be implemented for each The risk analysis to be implemented for each hazard is described below: hazard is described below:

[] []

Seismic Risks: EPRI Alternative Seismic Risks: EPRI Alternative Seismic Approach for Tier 1 plants Seismic Approach for Tier 1 plants identified in EPRI Report 3002022453 identified in EPRI Report 3002017583 (Reference 5). Vistra OpCo will follow a with the additional considerations similar alternative seismic approach in discussed in Section 3.2.3 of this LAR.

the 10 CFR 50.69 categorization Vistra OpCo will follow a similar process for CPNPP as the approach alternative seismic approach in the 10 that was approved by the NRC staff for CFR 50.69 categorization process for Calvert Cliffs Nuclear Power Plant CPNPP as the approach that was (ADAMS Accession No. approved by the NRC staff for Calvert ML19330D909) (Reference 19). Cliffs Nuclear Power Plant (ADAMS Accession No. ML19330D909)

(Reference 19). Calvert Cliffs Nuclear Power Plant supplementals responded to RAIs on the case studies identified in EPRI Report 3002012988. Updates to the EPRI Report 3002012988 from these case studies are identified in EPRI Report 3002017583.

to TXX-23094 Page 13 of 33

CPNPP LAR (Reference 36), Section 3.2.3 - Seismic Hazards is modified as shown below:

10 CFR 50.69(c)(1) requires the use of PRA to assess risk from internal events. For other risk hazards such as seismic, 10 CFR 50.69 (b)(2) allows, and NEI 00-04 summarizes, the use of other methods for determining SSC functional importance in the absence of a quantifiable PRA (such as Seismic Margin Analysis or IPEEE Screening) as part of an integrated, systematic process. For the CPNPP seismic hazard assessment, Vistra OpCo proposes to use a risk informed graded approach that meets the requirements of 10 CFR 50.69 (b)(2) as an alternative to those listed in NEI 00-04 sections 1.5 and 5.3. This approach is specified in EPRI Report 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," and includes additional qualitative considerations that are discussed in this section., where CPNPP meets the EPRI Report 3002017583 Tier 1 criteria for a Low Seismic Hazard/High Seismic Margin site. The Tier 1 criteria are as follows:

Tier 1: Plants where the GMRS [Ground Motion Response Spectrum] peak acceleration is at or below approximately 0.2g or where the GMRS is below or approximately equal to the SSE [Safe Shutdown Earthquake] between 1.0 Hz and 10 Hz. Examples are shown in Figures 2-1 and 2-2.

At these sites, the GMRS is either very low or within the range of the SSE such that unique seismic categorization insights are not expected.

Note: EPRI Report 3002017583 applies to the Tier 1 sites in its entirety except for sections 2.3 (Tier 2 sites), 2.4 (Tier 3 sites), Appendix A (seismic correlation), and Appendix B (criteria for capacity-based screening).

The Tier 1 criterion (i.e., basis) in EPRI Report 3002017583 is a comparison of the ground motion response spectrum (GMRS, derived from the seismic hazard) to the safe shutdown earthquake (SSE, i.e., seismic design basis capability). U. S. nuclear power plants that utilize the 10 CFR 50.69 Seismic Alternative (EPRI Report 3002017583) will continue to compare GMRS to SSE.

The trial studies in EPRI Report 3002017583 show that seismic categorization insights are overlaid by other risk insights even at plants where the GMRS is far beyond the seismic design basis. Therefore, the basis for the Tier 1 classification and resulting criteria is not that the design basis insights are adequate. Instead, it is that consideration of the full range of the seismic hazard produces limited unique insights to the categorization process. That is the basis for the following statements in Table 4-1 of the EPRI report.

"At Tier 1 sites, the likelihood of identifying a unique seismic condition that would cause an SSC to be designated HSS is very low.

Therefore, with little to no anticipated unique seismic insights, the 50.69 categorization process using the FPIE PRA and other risk evaluations along with the required Defense-in-Depth and IDP qualitative considerations are expected to adequately identify the safety-significant functions and SSCs required for those functions and no additional seismic reviews are necessary for 50.69 categorization."

This The proposed categorization approach for CPNPP is a risk-informed graded approach that is demonstrated to produce categorization insights equivalent to a seismic PRA. For Tier 1 plants, this approach relies on the insights gained from the seismic PRAs examined in EPRI Report 3002017583 along with confirmation that the site GMRS is low. EPRI Report 3002017583 demonstrates that seismic risk is adequately addressed for Tier 1 sites by the results of additional qualitative assessments discussed in this section and existing elements of the 50.69 categorization process specified in NEI 00-04.

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For example, the 50.69 categorization process as defined in NEI 00-04 includes an Integral Assessment that weighs the hazard-specific relative importance of a component (e.g., internal events, internal fire, seismic) by the fraction of the total Core Damage Frequency (CDF) contributed by that hazard. The risk from an external hazard can be reduced from the default condition of HSS if the results of the integral a ssessment meet the importance measure criteria for LSS. For Tier 1 sites, the seismic risk (CDF/LERF) will be low such that seismic hazard risk is unlikely to influence an HSS decision. In applying the EPRI Report 3002017583 process for Tier 1 sites to the CPNPP 10 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the EPRI Report 3002017583 guidance and informed of plant SSC-specific seismic insights for their consideration in the final HSS/LSS determinations.

EPRI Report 3002017583 recommends a risk-informed graded approach for addressing the seismic hazard in the 50.69 categorization process. There are a number of seismic fragility fundamental concepts that support a graded approach and there are important characteristics about the comparison of the seismic design basis (represented by the SSE) to the site-specific seismic hazard (represented by the GMRS) that support the selected thresholds between the three evaluation Tiers in the EPRI report. The coupling of these concepts with the categorization process in NEI 00-04 are the key elements of the approach defined in EPRI Report 3002017583 for identifying unique seismic insights.

The seismic fragility of an SSC is a function of the margin between an SSCs seismic capacity and the site-specific seismic demand. References such as EPRI NP-6041 (Reference 18) provide inherent seismic capacities for most SSCs that are not directly related to the site-specific seismic demand. This inherent seismic capacity is based on the non-seismic design loads (pressure, thermal, dead weight, etc.) and the required functions for the SSC. For example, a pump has a relatively high inherent seismic capacity based on its design and that same seismic capacity applies at a site with a very low demand and at a site with a very high demand. At sites with lower seismic demands such as CPNPP, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference

18. Low seismic demand sites have lower likelihood of seismically-induced failures and lesser challenges to plant systems. This, therefore, provides the technical basis for allowing use of a graded approach for addressing seismic hazards at CPNPP.

There are some plant features such as equipment anchorage that have seismic capacities more closely associated with the site-specific seismic demand since those specific features are specifically designed to meet that demand. However, even for these features, the design basis criteria have intended conservatisms that result in significant seismic margins within SSCs.

These conservatisms are reflected in key aspects of the seismic design process. The SSCs used in nuclear power plants are intentionally designed using conservative methods and criteria to ensure that they have margins well above the required design bases. Experience has shown that design practices result in margins to realistic seismic capacities of 1.5 or more.

The following provides the basis for establishing Tier 1 criteria in EPRI Report 3002017583:

a. SSCs for which the inherent seismic capacities are applicable, or which are designed to the plant SSE will have low probabilities of failure at sites where the peak spectral acceleration of the GMRS < 0.2g or where the GMRS < SSE between 1 and 10 Hz.

b. The low probabilities of failure of individual components would also apply to components considered to have correlated seismic failures.

c. These low probabilities of failure lead to low seismic CDF and LERF estimates, from an absolute risk perspective. to TXX-23094 Page 15 of 33

d. The low seismic CDF and LERF estimates lead to reasonable confidence that seismic risk contributions would allow reducing a HSS to LSS due to the 50.69 Integral Assessment if the equipment is HSS only due to seismic considerations.

Test cases described in Section 3 of EPRI Report 3002017583 showed that it would be unusual even for moderate hazard plants to exhibit any unique seismic insights, including due to correlated failures. The plant specific EPRI Report 3002017583 test case information Vistra OpCo is using from other licensees and being incorporated by Reference into this application is described in Case Study A (Reference 29, Reference 50, Reference 51), Case Study C (Reference 30, Reference 31), and Case Study D (Reference 32, Reference 33, Reference 34, Reference 52, Reference 53). Hence, while it is prudent to perform additional evaluations to identify conditions where correlated failures may occur for Tier 2 sites, for Tier 1 sites such as CPNPP, correlation studies would not lead to new seismic insights or affect the baseline seismic CDF in any significant way.

The Tier 1 to Tier 2 threshold as defined in EPRI Report 3002017583 provides a clear and traceable boundary that can be consistently applied, plant site to plant site. Additionally, because the boundary is well defined, if new information is obtained on the site hazard, a sites location within a particular Tier can be readily confirmed. In the unlikely event that the CPNPP seismic hazard changes to medium risk (i.e., Tier 2) at some future time, Vistra OpCo will follow its categorization review and adjustment process procedures to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e).

As discussed in subsection Evaluation of Seismic Hazard in Section 3.2.4.1.3 of Reference 1, Vistra OpCos approach for including the seismic risk contribution in the RICT calculation was to add a penalty seismic CDF and a penalty seismic LERF to each RICT calculation. The proposed bounding seismic CDF estimate was based on using the plant-specific mean seismic hazard curve developed in response to the Near-Term Task Force Recommendation 2.1 (Reference 20), and a plant level mean high confidence of low probability of failure (HCLPF) capacity of 0.12g referenced to peak ground acceleration (PGA). The uncertainty parameter for seismic capacity was represented by a composite beta factor of 0.4. The calculated seismic CDF penalty is 1.93E-06 per year. Concerning the proposed bounding seismic LERF estimate, an estimate of the seismic LERF was obtained by convolving the estimated seismic CDF (as described above) with a limiting fragility for containment integrity, also assu med to be 0.12g PGA HCLPF because the containment fragility is not available. The calculated seismic LERF is 9.73E-07 per year. This is used to assess the total baseline risk of the PRA, including the seismic contribution, remains less than 1E-04 per year for CDF and 1E-05 per year for LERF.

The small percentage contribution of seismic to total plant risk makes it unlikely that an integral importance assessment for a component, as defined in NEI 00-04, would result in an overall HSS determination. Further, the low hazard relative to plant seismic capability makes it unlikely that any unique seismic condition would exist that would cause an SSC to be designated HSS for a Tier 1 site such as Comanche Peak.

The following provides the basis for concluding that CPNPP meets the Tier 1 site criteria. In response to the NRC 50.54(f) letter associated with post-Fukushima recommendations (Reference 22), Vistra OpCo submitted a seismic hazard screening report for CPNPP (Reference 23) to the NRC.

The maximum GMRS value for CPNPP in the 1-10 Hz range meets the Tier 1 criterion of approximately 0.2g in Reference 5 EPRI Report 3002017583.

The CPNPP SSE and GMRS curves from the seismic hazard and screening response are to TXX-23094 Page 16 of 33

shown in Reference 23, Figure 4.1-1 (Figure 3-2 below). The NRCs staff assessment of the CPNPP seismic hazard and screening response is documented in Reference 24. In Section 3.4 of Reference 24 the NRC concluded that the methodology used by Vistra OpCo in determining the GMRS was acceptable and that the GMRS determined by Vistra OpCo adequately characterizes the reevaluated hazard for the CPNPP site.

Figure 3-2: Reference 23, Figure 4.1-1

Section 1.1.3 of EPRI Report 3002017583 cites various post-Fukushima seismic reviews performed for the U.S. fleet of nuclear power plants. For CPNPP, the specific seismic reviews prepared by the licensee and the NRCs staff a ssessments are provided here. These licensee documents were submitted under oath and affirmation to the NRC.

1. NTTF Recommendation 2.1 seismic hazard screening (Reference 23, Reference 24)

2. NTTF Recommendation 2.3 seismic walkdowns (Reference 25, Reference 26, Reference 27, Reference 28)

3. NTTF Recommendation 4.2 seismic mitigation strategy assessment (S-MSA) (Reference 48, Reference 49)

As an enhancement to the EPRI study results as they pertain to CPNPP, the proposed CPNPP categorization approach for seismic hazards will include qualitative consideration of the mitigation capabilities of SSCs during seismically-induced events and seismic failure modes, based on insights obtained from prior seismic evaluations performed for CPNPP. For example, as part of the categorization team's preparation of the System Ca tegorization Document (SCD) that is presented to the IDP, a section will be included in the SCD that summarizes the identified plant seismic insights pertinent to the system being categorized and will also state the basis for applicability of the EPRI Report 3002017583 study and the bases for CPNPP being a Tier 1 plant. The discussion of the Tier 1 bases will include such factors as:

The low seismic hazard for the plant, which is subject to periodic reconsideration as new information becomes available through industry evaluations; and to TXX-23094 Page 17 of 33

The definition of Tier 1 in the EPRI study.

At several steps of the categorization process (e.g., as noted in Table 3-1) the categorization team will consider the available seismic insights rela tive to the system being categorized and document their conclusions in the SCD. Integrated importance measures over all modeled hazards (i.e.,

internal events, including internal flooding, and internal fire for CPNPP) are calculated per Section 5.6 of NEI 00-04, and components for which these measures exceed the specified criteria are preliminary HSS which cannot be changed to LSS.

For HSS SSCs uniquely identified by the CPNPP PRA models but having design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events, these will be addressed using non-PRA based qualitative assessments in conjunction with any seismic insights provided by the PRA.

For components that are HSS due to fire PRA but not HSS due to internal events PRA, the categorization team will review design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events and characterize those functions for presentation to the IDP as additional qualitative inputs, which will also be described in the SCD.

The categorization team will review available CPNPP plant-specific seismic reviews and other resources such as those identified above. The objecti ve is to identify plant-specific seismic insights derived from the above sources, relevant to the components in the system being categorized, that might include potentially important impacts such as:

Impact of relay chatter Implications related to potential seismic interactions such as with block walls Seismic failures of passive SSCs such as tanks and heat exchangers Any known structural or anchorage issues with a particular SSC Components that are implicitly part of PRA-modeled functions (including relays)

Components that may be subject to correlated failures

Such impacts would be compiled on an SSC basis. As each system is categorized, the system-specific seismic insights will be provided to the IDP for consideration as part of the IDP review process. As such, the IDP can challenge, from a seismic perspective, any candidate LSS recommendation for any SSC if they believe there is basis for doing so. Any decision by the IDP to downgrade preliminary HSS components to LSS will also consider the applicable seismic insights in that decision. These insights will provide the IDP a means to consider potential impacts of seismic events in the categorization process.

Use of the EPRI approach outlined in Reference 21 to assess seismic hazard risk for 50.69 with the additional reviews discussed above will provide a process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs that satisfies the requirements of 10 CFR 50.69(c).

Based on the above, the Summary/Conclusion/Recommendation from Section 2.2.3 of EPRI Report 3002017583 applies to CPNPP, i.e., CPNPP is a Tier 1 plant for which the GMRS is very low such that unique seismic categorization insights are expected to be minimal. As discussed in EPRI Report 3002017583, the likelihood of identifying a unique seismic insight that would cause an SSC to be designated HSS is very low. Therefore, with little to no anticipated unique seismic insights, the 50.69 categorization process using the Full Power Internal Events (FPIE) PRA and other risk evaluations along with the defense-in-depth and qualitative assessment by the IDP adequately identifies the safety-significant functions and SSCs. to TXX-23094 Page 18 of 33

Feedback and Adjustment Process Impacts

To address the feedback and adjustment (i.e., performance monitoring) process as it pertains to the proposed CPNPP Tier 1 approach discussed above, implementation of the Vistra OpCo design control and corrective action programs will ensure the inputs for the qualitative determinations for seismic continue to remain valid to maintain compliance with the requirements of 10 CFR 50.69(e).

The performance monitoring process is described in Vistra OpCo 10 CFR 50.69 program documents. The program requires that the periodic review assess changes that could impact the categorization results and provides the IDP with an opportunity to recommend categorization and treatment adjustments. Personnel from Engineer ing, Operations, Risk Management, Regulatory Affairs, and others have responsibilities for preparing and conducting various performance monitoring tasks that feed into the performance monitoring process. The intent of the performance monitoring reviews is to discover trends in component reliability, to help catch and reverse negative performance trends and to take corrective action if necessary.

The Vistra OpCo configuration control process ensures that changes to the plant, including a physical change to the plant and changes to documents, are evaluated to determine the impact to drawings, design bases, licensing documents, programs, procedures, and training.

Vistra OpCo has a comprehensive problem identification and corrective action program that ensures that issues are identified and resolved. Any issue that may impact the 10 CFR 50.69 categorization process will be identified and addressed through the problem identification and corrective action program, which would encompass seismic-related issues.

The Vistra OpCo 10 CFR 50.69 program requires that System Categorization Documents cannot be approved by the IDP until the panel's comments have been resolved to the satisfaction of the IDP. This includes issues related to system-specific seismic insights considered by the IDP during categorization.

All other aspects of the CPNPP 50.69 Feedback and Review process remain as stated in this LAR as identified in Section 3.5.

The Periodic Review process, as described else where in the LAR, assesses system/component performance changes and plant operation or design changes that have occurred for categorized systems to review the impact of plant changes on RISC-1, RISC-2, RISC-3, and RISC-4 SSCs.

The review will be implemented by station procedure.

The Vistra OpCo design process includes, but is not limited to, the following:

Requirements for redundancy, diversity, and separation of structures, system and components are met, including seismic interactions Review of impact to seismic loading, SSE seismic requirements Review of seismic dynamic qualification of components if the configuration change adds, relocates, or alters Seismic Category I mechanical or electrical components

CPNPP LAR (Reference 36), Section 3.5 - Feedback and Adjustment Process will have the addition of this paragraph at the end of the section:

The periodic monitoring requirements of the 10 CFR 50.69 process will ensure that these issues are captured and addressed at a frequency commensurate with the issue significance. The periodic monitoring program includes immediate and periodic reviews, that include the to TXX-23094 Page 19 of 33

requirements of the regulation, to ensure that all issues that could affect 10 CFR 50.69 categorization are addressed. The periodic monitori ng process also monitors the performance and condition of categorized SSCs to ensure that the assumptions for reliability in the categorization process are maintained.

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b) If differences exist between the licensees proposed alternate seismic approach and the information in the supplements stated above, identify such differences and either incorporate them in the proposed approach or justify their exclusion.

Response

The following provides a high-level discussion on major differences between the supplements stated above compared to the CPNPP approach. Note that minor and editorial differences may not be mentioned in this list. To provide for a comprehensive examination of the 10 CFR 50.69 license amendment request and CPNPP proposed modificati ons to the License Amendment Request are shown in the response to APLC 01.a.

Supplement to the Calvert Cliffs 10 CFR 50.69 LAR dated May 10, 2019 (Reference 38)

The Calvert Cliffs Supplement modified the te xt in the LAR which are described below:

Section 3.1.1 This supplement at Calvert Cliffs modified Section 3.1.1 to add additional information on seismic. CPNPP has added similar discussion.

Figure 3-1 shown in the Calvert Cliffs supplemental is not present in the CPNPP LAR.

The Seismic Risk bullet in Section 3.1.1 has been modified but differs from the text in the Calvert Cliff supplemental so as to provide information on the linkage to the plant case studies of the EPRI Report evaluated in the Calvert Cliffs supplements.

Section 3.2.3 This supplement at Calvert Cliffs modified Section 3.2.3 to add additional information on seismic. CPNPP has added similar discussion.

References of test cases have been added to be consistent with more recent LAR submittals.

The paragraph starting with Use of the EPRI approach has been modified for additional clarity.

The following two discussions were not included in CPNPP:

o The site-specific Calvert Cliffs information (e.g., seismic capacity discussions, etc.) from the 50.69 and other Calvert Cliffs licensing responses do not apply to CPNPP. CPNPP-specific seismic capacity information is described in Section 3.2.3. The categorization team will provide the IDP with additional insights from previous seismic evaluations at CP NPP, as described in this LAR.

o The configuration control checklist described in the Calvert Cliffs response implies that a specific checklist was developed for 50.69 reviews. Refer to the discussion above for the configuration control and periodic review processes.

Section 3.5 Similar wording to Calvert Cliff was included in: (1) the CPNPP Section 3.2.3 modifications for seismic-related discussions, and (2) a new paragraph was added to the end of Section 3.5.

Supplement to the Calvert Cliffs 10 CFR 50.69 LAR dated July 1, 2019 (Reference 39)

The Calvert Cliffs Supplement RAI 4.a and RAI 4.b discussed seismic.

RAI 4 further clarifies the information related to the alternate seismic approach and is applicable to CPNPP.

RAI 5, 6, and 8 do not correspond to seismic evaluations.

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Supplement to the Calvert Cliffs 10 CFR 50.69 LAR dated July 19, 2019 (Reference 40)

The plant case studies identified and the responses to this in RAIs 1, 2, and 3 are generic to the overall EPRI report evaluation. This Calvert Cliffs Nuclear Power Plant supplement responded to RAIs on the case studies identified in EPRI Report 3002012988. These case studies are generic in order to evaluate the process of the EPRI report and therefore are valid for CPNPP. Updates to the EPRI Report 3002012988 from these case studies are identified in EPRI Report 3002017583. RAI 7 is not associated with seismic.

Supplement to the Calvert Cliffs 10 CFR 50.69 LAR dated August 5, 2019 (Reference 41)

This supplement revised the response to RAI 3 from the Supplement to the Calvert Cliffs 10 CFR 50.69 LAR dated July 19, 2019 (Reference 40). These case studies are generic in order to evaluate the process of the EPRI report and therefore are valid for CPNPP.

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Section 3.2.3 of the LAR stated that the following provides the basis for concluding that Comanche Peak meets the Tier 1 site criteria:

The maximum GMRS [ground motion response s pectrum] value for CPNPP in the 1-10 Hz range meets the Tier 1 criterion of approximately 0.2g in Reference 5.

However, the LAR did not provide GMRS and safe shutdown earthquake (SSE) curves, or cited it from a docketed report.

c) Provide an SSE/GRMS hazard curve comparis on or cite it from a reference to demonstrate that the Comanche Peak Tier 1 criterion is met for the applicable frequency band.

Response

The following information is shown below and is in the proposed modifications to Section 3.2.3 shown in the response to APLC 01.a.

The CPNPP SSE and GMRS curves from the seismic hazard and screening response are shown in Reference 23, Figure 4.1-1 (Figure 3-2 below). The NRCs staff assessment of the CPNPP seismic hazard and screening response is documented in Reference 24. In Section 3.4 of Reference 24 the NRC concluded that the methodology used by Vistra OpCo in determining the GMRS was acceptable and that the GMRS determined by Vistra OpCo adequately characterizes the reevaluated hazard for the CPNPP site.

Figure 3-2: Reference 23, Figure 4.1-1

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APLC Question 02 - Development of the Tornado Safe Shutdown Equipment List (TSSEL)

Paragraph 50.69(b)(2)(ii) of 10 CFR requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs. Section 3.3 of NEI 00-04, Revision 0 provides limited guidance for determining the technical adequacy attributes required for these types of analyses for this specific application. RG 1.201, Revision 1 states that as part of the plant-specific application requesting to implement §50.69, the licensee or applicant will provide the bases supporting the technical adequacy of itsnon-PRA-type analyses for this application.

Section 3.2.4 of the LAR stated that the tornado hazard safety significance process uses a TSSEL that was developed. The LAR stated that the proposed approach will identify SSCs associated with this hazards safe shutdown function and barriers to be high safe significant (HSS). However, the licensee didnt discuss how the list was developed in the LAR. It is also not clear how the TSSEL will be managed to ensure t he impact of plant and procedure changes that could impact the TSSEL are addressed in CFR 50.69 categorization.

a) Explain how the TSSEL was developed. Provide details on the methodology to develop the TSSEL. Include in the response how the high wind related SSCs would be processed using this methodology.

Response

Upon further research, it was determined that CPNPP does not currently have a formal Tornado Safe Shutdown Equipment List. CPNPP does have lists of components with design functions related to the tornado and straights winds hazards (collectively referred to in these responses as high winds), but these lists were not evaluated or arranged in the format of a SSEL.

For the 10CFR50.69 program, an explicit High Wi nds Safe Shutdown Equipment Lists (HWSSEL) will be developed for use during the categorizatio n process. This SSEL will encompass hazards related to tornados and straight winds, which will be referred to, in the collective, as High Winds.

The HWSSEL will be developed using a process similar to that used to develop the seismic SSEL.

The HWSSEL will be developed prior to performing any system categorization.

When the seismic SSEL was originally created at CPNPP, the internal events PRA model was not peer reviewed under the current standard. The seismic SSEL was originally developed addressing the safe shutdown functions of (1) Decay Heat Removal, (2) Reactivity Control, (3) Inventory Control, (4) Power Availability, and (5) Reactor Pressure Control. Now, CPNPP has an internal events PRA model that is peer reviewed under the current standard and that has properly assessed the equipment needed to support the same safety functions, therefore the internal events PRA can be used to inform the identification of SSCs of the HWSSEL. The impact of high winds are addressed by the HWSSEL development process as described below. The following steps will be undertaken to complete the HWSSEL:

1. From the internal events PRA model, identify safe shutdown equipment resulting from initiating events that can be caused by tornadoes or straight winds including missiles.

2. From this list of initiating events, an initial HWSSEL will be selected from the structure of the internal events PRA model. This step will be completed by selecting basic events from the PRA model that correspond to SSCs whose functionality would prevent core damage and large early release. This is similar to the concept of how the seismic SSEL was created which identified SSCs required to achieve and maintain safe shutdown of the plant. to TXX-23094 Page 24 of 33

3. From this list of SSCs, SSCs can be eliminat ed if they do not result in failing a safety function.

4. This list of SSCs will be reviewed and augmented as necessary to include SSCs relied upon to maintain the tornado design criteria, such as doors, vents, dampers, and louvers if those types of SSCs are being categorized in a system categorization.

5. The resulting list will be the HWSSEL for the plant.

b) Provide justification that the TSSEL method meets the expectations in the Statements of Consideration for 10 CFR 50.69 that non-PRA methods used in the categorization process are conservative. In the justification identify the industry assessments referenced in the LAR and summarize the industry evaluations and results that support the conclusion that the Comanche Peaks proposed approach to use the TSSEL is sufficient to support a robust 50.69 categorization process.

Response

The HWSSEL method is conservative since it identifies SSCs that support the safe shutdown pathway to be required to be HSS with its process discussed in the response to APLC 02.a. A HWSSEL (or a Tornado SSEL) has been previously identi fied in Arkansas Nuclear One, Unit 1 (Reference 42) and Unit 2 (Reference 43) for use in 10 CFR 50.69 categorization which was approved in Reference 44 for Arkansas Nuclear One, Unit 1 and Reference 45 (with correction in Reference 46) for Arkansas Nuclear One, Unit 2. Identification of SSCs on the HWSSEL is a bounding assessment since it does not take into account the probability of the SSC successfully providing a safe shutdown pathway. This is consistent with the concept of the seismic margins analysis and Appendix R Safe Shutdown analysis bei ng able to be used rather than PRA models for the seismic and fire hazards respectively. This matches the concept as stated in the SOC (Reference 47):

The supporting guidance for the rule has been structured such that the licensees will gain more benefit when PRA methods are used (beyond the minimum PRA requirements in § 50.69(c)), and where non-PRA methods are used, the requirements and associated implementation guidance account for this situat ion by requiring a process that tends to conservatively categorize SSCs into RISC-1 and RISC-2 (i.e., no special treatment requirements removed).

The HWSSEL will not be screening SSCs based on their probability of failure or occurrence of the initiating event from straight winds or tornados and instead will be a conservative evaluation of SSCs required for safe shutdown of CPNPP. Theref ore, this meets the requirements of the SOC and is consistent with other industry approaches for other hazards that have been evaluated with qualitative considerations rather than PRA models.

c) Explain how the non-tornado high wind hazard is encompassed by the proposed approach.

Response

The non-tornado (straight wind) hazard would be encompassed in the HWSSEL based on the methodology identifying SSCs that would be required to be on the safe shutdown equipment list in the scenario an initiating event occurs that could be caused by a tornado or straight wind event.

Steps 1, 3, and 4 of the process described in the response to APLC 02.a of this question involve contribution from the tornado or straight wind hazard. The process encompasses both hazards because the potential initiating events that can be caused by a tornado bound those that can be caused by a straight wind event. Likewise, the tornado design criteria are bounding relative to the plausible maximum straight wind event (which would not include a substantial atmospheric pressure change and would have a lower maximum wind speed than the tornado design criteria). to TXX-23094 Page 25 of 33

d) Confirm that the TSSEL is an active document that reflects the current as-built, as-operated plant and that changes to the plant will be evaluated to determine their impact to the equipment list and the categorization process.

Response

The HWSSEL will be a document that is created for the 10 CFR 50.69 categorization. As part of the periodic assessment of systems categorized, it will be confirmed that the TSSEL has not been impacted by plant changes and still reflects the current as-built, as-operated plant.

e) Identify and explain of the program that will manage the TSSEL.

Response

As part of the CPNPP PRA configuration control process, changes to the plant design and operations are periodically reviewed for potential impact to the PRA model and the approved risk-informed applications. Upon approval of the 10 CFR 50.69 application, this periodic review will include all aspects of the 50.69 program, including the HWSSEL input to the categorization.

In addition, in Section 3.2.4 of the LAR, the licensee discussed the equipment that fulfills a TSSEL function. Table 3-1, Categorization Evaluation Summary, includes an item Fire and Other External Hazards, and high winds and tornado hazards are not specifically mentioned. If the high winds and tornados are considered as a part of Fire and Other External Hazards, its evaluation level is for component only, which may not include functions from the TSSEL, which are discussed in Section 3.2.4 of the LAR.

f) Table 3-1, Categorization Evaluation Summary, lists the different methods for categorization and the rules under which the IDP can or cannot change HSS to low safety significant (LSS). The high winds initiator is not listed separately in this Table.

Because high winds are evaluated differently than all other external hazards at Comanche Peak, explain why this hazard is not specifically provided in the table or revise the table to provide high winds s eparately to show that candidate TSSEL SSCs cannot be changed by the IDP and that the evaluation level can be at the function level as well.

Response

Table 3-1 is proposed to be modified as shown below; refer to the footnote 3 for information on the HWSSEL function examination.

Table 3-1. Categorization Evaluation Summary

Categorization IDP Change HSS Drives Element Step - NEI 00-04 Evaluation Level to LSS Associated Section Functions Risk (PRA Internal Events Modeled) Base Case - Component Not Allowed Yes Section 5.1 to TXX-23094 Page 26 of 33

Categorization IDP Change HSS Drives Element Step - NEI 00-04 Evaluation Level to LSS Associated Section Functions Fire, Seismic and other External Allowable No Events Base Case PRA Sensitivity Studies Allowable No Integral PRA Assessment - Not Allowed Yes Section 5.6 Fire and Other Allowed No External Hazards Component Not Tornado and Risk (Non-Straight Winds Component(3) Not Allowed No modeled) Seismic -

Alternative Tier 1 Function/Component Allowed(1) No Approach Shutdown - Allowed No Section 5.5 Function/Component Not Core Damage - Allowed Yes Defense-in-Section 6.1 Function/Component Not Depth Containment - Allowed Yes Section 6.2 Component Not Qualitative Considerations -

Criteria Section 9.2 Function Allowable(2) N/A

Passive Passive - Section 4 Segment/Component Not Allowed No

Notes:

(1) IDP consideration of seismic insights can also result in an LSS to HSS determination.

(2) The assessments of the qualitative considerations are agreed upon by the IDP in accordance with NEI 00-04, Section 9.2. In some cases, a 50.69 categorization team may provide preliminary assessments of the seven considerations for the IDPs consideration, however the final assessments of the seven considerations are the direct responsibility of the IDP.

The seven considerations are addressed preliminarily by the 50.69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step. Each of the seven considerations requires a supporting justification for confirming (true response) or not confirming (false response) that consideration. If the 50.69 categorization team determines that one or more of the seven considerations cannot be confirmed, then that function is presented to the IDP as preliminary HSS. Conversely, if all the seven considerations are confirmed, then the function is presented to the IDP as preliminary LSS.

The System Categorization Document, including the justifications provided for the qualitative considerations, is reviewed by the IDP. The IDP is responsible for reviewing the preliminary assessment to the same level of detail as the 50.69 team (i.e., all considerations for all functions are reviewed). The IDP may confirm the preliminary function risk and associated justification or may direct

to TXX-23094 Page 27 of 33

that it be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the IDP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the IDP. If the IDP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS.

(3) Refer to Figure 3-1 and Section 3.2.4 for more information on the evaluation level. This evaluation does examine if a component supports a HWSSEL function which is evaluated for each component.

to TXX-23094 Page 28 of 33

APLC Question 03 - External Flooding

Paragraph 50.69(b)(2)(ii) of 10 CFR requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs.

Table 2 of the LAR supplement dated June 8, 2023, provided the External Hazard Screening. In the external flooding section, it states Effects of local intense precipitation (LIP) are incorporated into the Probable Maximum Flood (PMF) for the site, which is used in bounding external flooding calculations. Subsequent engineering evaluation determined resulting flood heights within safety-related structures did not affect mitigating strategy equipment and were bounded by internal flood results. However, the licensee did not provide any SSCs that are credited to screen out the external flooding, such as exterior doors.

Provide a list of the specific exterior doors or any SSCs that will be assigned HSS, since they are credited for screening the external flood hazard in accordance with Figure 5-6 in NEI 00-04.

Response

SSCs credited for screening for all other external hazards (excludes seismic which is addressed in Section 3.2.3 of the CPNPP LAR, and high winds and tornadoes which are addressed in its own subsection in Section 3.2.4 of the CPNPP LAR) will be evaluated using the guidance illustrated in Figure 5-6 of NEI 00-04 during the implementation of the 10 CFR 50.69 categorization process. The syst em categorization process requires that every SSC within the system is evaluated with respect to Figure 5-6 of NEI 00-04 for all other external hazards including the consideration if failure to credit the component results in an unscreened scenario (e.g., exterior doors that would be credited for the screening of external flooding). Since this rigorous categorization process examines all SSCs within the system for evaluation of all other external hazards, specific exterior doors will be identified if they are within a system categorized. During system categorization, the categorization team has detailed knowledge on the workings of the system and the evaluation requirements which allows for a detailed consideration of the external flooding hazard in accordance with Figure 5-6 of NEI 00-04. Note that it is unlikely that CPNPP will have every system categorized within the plant. Therefore, every exterior door that would impact the external flood hazard may not be categorized; but if they are in a system being categorized, these external doors (and all other SSCs within the system) would undergo the rigorous evaluation during system categorization.

In response to the specific request concerning exterior doors or other SSCs credited for screening the external flood hazard, CPNPP has no doors or other components that would be credited in the screening. The plants original design and the reevaluation in response to the 50.54(f) Information request Regarding Near-Term Task Force Recommendation 2.1:

Flooding (Reference 54), concluded that there are no specific doors credited for protection of safety related components due to external flooding concerns. The doors on the west face of the Unit 2 Turbine Building (TB), elevation 809.3 ft, are not required to maintain their function due to the availability of the Unit 2 TB sump/condenser pit to capture flood waters due to participation/runoff. Because of this design feature, the three doorways that provide entrance to the safety related Electrical Control Building from the Turbine Building at elevation 778 ft, are also not required to function from this external flood hazard.

For the actual categorization, the evaluations referenced in the NTTF 2.1 response would be appropriately reviewed, assessed, and documented. to TXX-23094 Page 29 of 33

References The reference list that was identified in the CPNPP LAR (Reference 36) is used as the numbering format. No references have been removed even if they are no longer used in the CPNPP proposed 10 CFR 50.69 process (e.g., EPRI 3002022453).

1. Letter from the NRC to Vistra OpCo Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 183 and 183 Regarding the Adoption of Technical Specification Task Force Traveler TSTF-505, Revision 2 (EPID L-2021-LLA-0085), August 22, 2022 (ADAMS Accession No. ML22192A007).

2. NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, Nuclear Energy Institute, July 2005.

3. NRC Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, May 2006.

4. NUREG/CR-6850 (also EPRI 1011989), "Fire PRA Methodology for Nuclear Power Facilities," September 2005, with Supplement 1 (EPRI 1019259), Fire Probabilistic Risk Assessment Methods Enhancements, September 2010.

5. Electric Power Research Institute TR 3002022453, Alternate Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, 2021.

6. NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management, December 1991.

7. Letter from the NRC to Entergy Operations Inc, Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative AN02-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replac ement Activities in Class 2 and 3 Moderate and High Energy Systems (TAC NO. MD5250), April 22, 2009, (ML090930246).

8. Letter from the NRC to Southern Nuclear O perating Company, Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC Nos.

ME9472 and ME9473), December 17, 2014, (ML14237A034).

9. NRC, RG 1.147, Rev. 15, Inservice Inspec tion Code Case Acceptability, ASME Section XI, Division 1, October 2007.

10. Letter from Vistra OpCo to the NRC, Applicatio n to Revise Technical Specifications to Adopt Risk Informed Completion Times, TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b', May 11, 2021, (ML21131A233).

11. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009.

12. NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, March 2017.

13. EPRI TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, December 2008. to TXX-23094 Page 30 of 33

14. NRC Regulatory Issue Summary 2007-06, Regul atory Guide 1.200 Implementation, March 22, 2007.

15. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018.

16. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Appl ications, Addendum A to RA-S-2008, February 2009.

17. Letter from Vistra OpCo to the NRC, License Amendment Request (LAR)20-006 Application To Revise Technical Specifications To Adopt Risk Informed Completion Times, TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b Request For Additional Information (RAI), March 29, 2022, (ML22088A299).

18. Electric Power Research Institute NP-6041-SL, A Methodology for Assessment of Nuclear Plant Seismic Margin, Revision 1, August 1991.

19. Letter from the NRC to Exelon Generation Company, LLC, Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 332 and 310 RE: Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2018-LLA-0482), February 28, 2020, (ML19330D909).

20. Letter from the NRC to All Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," March 12, 2012, (ML12073A348).

21. Electric Power Research Institute Report 3002017583, Alternative Approaches for Addressing Seismic Risk in 10CFR50.69 Risk-Informed Categorization, July 2018, (ML21082A170).

22. Letter from the NRC to All Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1,2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, March 12, 2012, (ML12053A340).

23. Letter from Luminant Power to the NRC, Comanche Peak Nuclear Power Plant, Docket Nos. 50-445 AND 50-446, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 27, 2014, (ML14099A197).

24. Letter from the NRC to Luminant Power, Comanche Peak Nuclear Power Plant, Units 1 and 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(F), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident and Staff Closure of Activites Associated with Recommendation 2.1, Seismic (CAC Nos.

MF3937 and MF3938), January 22, 2016, (ML16014A125). to TXX-23094 Page 31 of 33

25. Letter from Luminant Power to the NRC, Comanche Peak Nuclear Power Plant, Docket Nos. 50-445 and 50-446, 120-Day Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, November 27, 2012, (ML13009A269).

26. Letter from the NRC to Luminant Power, Comanche Peak Nuclear Power Plant, Units 1 and 2 - Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident (TAC Nos. MF0110 and MF0111), March 7, 2014 (ML14036A036).

27. Letter from Luminant Power to the NRC, Comanche Peak Nuclear Power Plant, Docket Nos. 50-445 and 50-446, Supplement to 180-Day Response to NRC Request for Information Pursuant to 10 CFR 50.54 (f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, July 1, 2013 (ML13192A179).

28. Letter from Luminant Power to the NRC, Comanche Peak Nuclear Power Plant (CPNPP)

Docket Nos. 50-445 And 50-446 Submittal of Requested Information Identified During Regulatory Audit Regarding Seismic Walkdowns At CPNPP, Units 1 And 2, To Support Implementation Of Near-Term Task Force Recommendation 2.3 Related To The Fukushima Dai-Ichi Nuclear Power Plant Accident (TAC Nos. MFO110 And MFO111), September 12, 2013 (ML13267A158).

29. Letter from Exelon Generation to the NRC for Peach Bottom Atomic Power Station, Units 2 and 3, Seismic Probabilistic Risk Assessm ent Report, "Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," August 28, 2018, (ML18240A065).

30. Letter from Southern Nuclear to the NRC for Vogtle Electric Generating Plant, Units 1 and 2, License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process, June 22, 2017, (ML17173A875).

31. Letter from the NRC to Southern Nuclear for Vogtle Electric Generating Plant, Units 1 and 2, "Issuance of Amendments Regarding Application of Seismic Probabilistic Risk Assessment Into the Previously Approved 10 CFR 50.69 Categorization Process (EPID L-2017-LLA-0248)," August 10, 2018, (ML18180A062).

32. Letter from Tennessee Valley Authority to the NRC, Seismic Probabilistic Risk Assessment for Watts Bar Nuclear Plant, Units 1 and 2, "Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident," June 30, 2017, (ML17181A485).

33. Letter from Tennessee Valley Authority to the NRC, Tennessee Valley Authority (TVA) -

Watts Bar Nuclear Plant Seismic Probabilistic Risk Assessment Supplemental Information, April 10, 2018, (ML18100A966).

34. Letter from Tennessee Valley Authority to the NRC, Watts Bar Nuclear Plant, 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 (WBN-TS-17-24),"

November 29, 2018, (ML18334A363). to TXX-23094 Page 32 of 33

35. Letter from PSEG to the NRC for Hope Creek Generating Station, Response to Request for Additional Information, Re: Adopt 10 CFR 50.69 LAR, June 25, 2020, (ML20177A535).

36. Letter from Vistra OpCo to the NRC for CPNPP, Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of st ructures, systems and components for nuclear power reactors. LAR 23-001, April 19, 2023, (ML23109A333).

37. Letter from Vistra OpCo to the NRC for CPNPP, Response to Request for Supplemental Information for License Amendment Request to Adopt 10 CFR 50.59, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2023-LLA-0057), June 8, 2023, (ML23159A200).

38. Letter from Exelon Generation to the NRC for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revised submittal to Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and com ponents for nuclear power reactors, May 10, 2019, (ML19130A180).

39. Letter from Exelon Generation to the NRC for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors, July 1, 2019, (ML19183A012).

40. Letter from Exelon Generation to the NRC for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors, July 19, 2019, (ML19200A216).

41. Letter from Exelon Generation to the NRC for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revised Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, Risk-Informed Categorization And Treatment of Structures, Systems, and Components for Nuclear Power Reactors, letter dated July 19, 2019, August 5, 2019, (ML19217A143).

42. Letter from Entergy to NRC for Arkansas Nuclear One, Unit 1, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, May 26, 2021, (ML21147A234).

43. Letter from Entergy to NRC for Arkansas Nuclear One, Unit 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, May 26, 2021, (ML21147A264).

44. Letter from NRC to Entergy, Arkansas Nuclear One, Unit 1 - Issuance of Amendment No.

277 RE: Adoption of 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2021-LLA-0105), June 23, 2022, (ML22138A431).

45. Letter from NRC to Entergy, Arkansas Nuclear One, Unit 2 - Issuance of Amendment No.

331 RE: Adoption of 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2021-LLA-0106), July 19, 2022, (ML22165A244). to TXX-23094 Page 33 of 33

46. Letter from NRC to Entergy, Arkansas Nuclear One, Unit 2 - Correction to Issuance of Amendment No. 331 RE: Adoption of 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (EPID L-2021-LLA-0106), August 26, 2022, (ML22236A545).

47. NRC, Federal Register, Volume 69, No. 224, 10 CFR Part 50 - RIN 3150-AG42 -

Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, November 22, 2004.

48. Letter from Luminant Power to the NRC for CPNPP, Mitigating Strategies Assessment (MSA) Report for the New Seismic Hazard Information NEI 12-06, Appendix H, Revision 2, H.4.1 Path 1: GMRS SSE, January 26, 2017, (ML17030A044).

49. Letter from NRC to Luminant Power for CPNPP, Comanche Peak Nuclear Power Plant, Units 1 and 2 - Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012 50.54(f) Letter (CAC Nos. MF7817 and MF7818), February 2, 2017, (ML17031A005).

50. Letter from NRC to Exelon Generation, Peach Bottom Atomic Power Station, Units 2 and 3 -

Staff Review of Seismic Probabilistic Risk Assessment, Associated with Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1:

Seismic, (EPID NO. L-2018-JLD-0010), June 10, 2019, (ML19053A469).

51. Letter from NRC to Exelon Generation, Peach Bottom Atomic Power Station, Units 2 and 3 -

Correction Regarding Staff Review of Seismic Probabilistic Risk Assessment, Associated with Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic," (EPID NO. L-2018-JLD-0010), October 8, 2019, (ML19248C756).

52. Letter from NRC to Tennessee Valley Authority, Watts Bar Nuclear Plant, Units 1 and 2 -

Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1:

Seismic (CAC NOS. MF9879 and MF9880; EPID L-2017-JLD-0044), July 10, 2018 (ML18115A138).

53. Letter from NRC to Tennesee Valley Authority, Watts Bar Nuclear Plant, Units 1 and 2 -

Issuance of Amendment Nos. 134 And 38 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 (EPID L-2018-LLA-0493),

April 30, 2020 (ML20076A194).

54. Luminant Generation Company LLCs Letter TXX-14094, Submittal of Fukushima Lessons Learned - Flood Hazard Reevaluation Report, Supplement 1, dated August 14, 2014 (ML14245A136) to TXX-23094 Pa ge 1 of 1

Revised License Condion

Vistra OpCo is approved to implement 10 CFR 50.69 using the processes for categorizaon of Risk-Informed Safety Class (RISC)-1, RISC -2, RISC -3, and RISC -4 structures, systems, and components (SSCs) using: Probabilisc Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal "ooding, and internal "re; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorizaon method to assess passive component risk for Class 2 and Class 3 and non-class SSCs and their associated supports; the results of the non-PRA evaluaons that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening signi"cance process iden"ed in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except wind-generated missiles and seism ic; the high winds safe shutdown equipment list for wind-generated missiles; and the alternave seismic approach as described in Vistra OpCos submital leter dated April 19, 2023, and all its subsequent associated supplements, as speci"ed in License Amendment No. [XXX] dated [DATE].

Vistra OpCo will complete the High Winds Safe Shutdown Equipment List (HWSSEL) prior to performing any system categorizaon per 10 CFR 50.69.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorizaon process speci"ed above (e.g., change from a seismic margins approach to a seismic probabilisc risk assessment approach).