CP-202300140, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. LAR 23-001

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Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. LAR 23-001
ML23109A333
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 04/19/2023
From: John Lloyd
Luminant, Vistra Operating Co. (VistraOpCo)
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
CP-202300140, TXX-23016, LAR 23-001
Download: ML23109A333 (1)
v • d • e


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-202300140 TXX-23016 April 19, 2023 ATTN: Document Control Desk Ref 10 CFR 50.90 U. S. Nuclear Regulatory Commission 10 CFR 50.91 Washington, DC 20555-0001 10 CFR 50.69

Subject:

Comanche Peak Nuclear Power Plant (CPNPP)

Docket Nos. 50-445 and 50-446 Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. LAR 23-001 Reference 1: Letter from NRC to Mr. Ken J. Peters, Issuance of Amendment Nos. 183 and 183 Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-505, Revision 2 (EPID L-2021-LLA-0085), August 22, 2022, ML22192A007

Dear Sir or Madam:

Pursuant to 10 CFR 50.69 and 10 CFR 50.90, Vistra Operations Company LLC (Vistra OpCo) is submitting a request for an amendment to the Operating Licenses for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (CPNPP). The proposed amendment would modify the CPNPP licensing bases, by the addition of License Conditions, to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation).

For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The enclosure to this letter provides the basis for the proposed changes to the CPNPP Operating Licenses.

The categorization process being implemented through this change is consistent with NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0 dated July 2005, with one exception discussed herein, which was endorsed by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, dated May 2006. To address seismic hazard in the Structures, Systems and Components (SSC) categorization process, an alternative method to NEI 00-04 has been implemented consistent with the Electric Power Research Institute (EPRI) Alternative Seismic Approach described in EPRI Report 3002022453. Attachment 1 of the enclosure provides a list of categorization prerequisites. Use of the categorization process on a plant system will only occur after these prerequisites are met.

TXX-23016 Page 2 of 2 The NRC has previously reviewed the technical adequacy of the CPNPP Probabilistic Risk Assessment (PRA) models for internal events, including internal flooding, and internal fire as described in Reference 1. Vistra OpCo requests the NRC utilize the review of the PRA technical adequacy for that application when performing the review for this application.

Vistra OpCo requests approval of the proposed license amendment within one year of completion of the NRC's acceptance review. The amendment will be implemented within 90 days after approval.

There are no new regulatory commitments made in this submittal.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of Texas Official.

Should you have any questions, please contact Nie Boehmisch at (254) 897-5064 or nicholas.boehmisch@luminant.com.

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

Executed on April 19, 2023.

Sincerely,

Enclosure:

Description and Assessment c (email) - Robert Lewis, Region IV [Robert.Lewis@nrc.gov]

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

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

David Nani, Resident Inspector, CPNPP [David.Nani@nrc.gov]

Mr. Robert Free [robert.free@dshs.state.tx.us]

Environmental Monitoring & Emergency Response Manager Texas Department of State Health Services Mail Code 1986 P. 0. Box 149347 Austin TX, 78714-9347

Enclosure to TXX-23016 Page 1 of 24 Evaluation of the Proposed Change 1

SUMMARY

DESCRIPTION 2 DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS 2.2 REASON FOR PROPOSED CHANGE

2.3 DESCRIPTION

OF THE PROPOSED CHANGE 3 TECHNICAL EVALUATION 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(B)(2)(I))

3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(B)(2)(II))

3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(B)(2)(III))

3.4 RISK EVALUATIONS (10 CFR 50.69(B)(2)(IV))

3.5 FEEDBACK AND ADJUSTMENT PROCESS 4 REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS

4.3 CONCLUSION

S 5 ENVIRONMENTAL CONSIDERATION 6 REFERENCES : List of Categorization Prerequisites : CPNPP Baseline PRA Model/Non-PRA Results

Enclosure to TXX-23016 Page 2 of 24 1

SUMMARY

DESCRIPTION The proposed amendment modifies the 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 (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

2 DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS The Nuclear Regulatory Commission (NRC) has established a set of regulatory requirements for commercial nuclear reactors to ensure that a reactor facility does not impose an undue risk to the health and safety of the public, thereby providing reasonable assurance of adequate protection to public health and safety. The current body of NRC regulations and their implementation are largely based on a deterministic approach.

This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBEs). The deterministic approach then requires that the facility include safety systems capable of preventing or mitigating the consequences of those DBEs to protect public health and safety. The Structures, Systems and Components (SSCs) necessary to defend against the DBEs are defined as safety-related, and these SSCs are the subject of many regulatory requirements, herein referred to as special treatments, designed to ensure that they are of high quality and high reliability and have the capability to perform during postulated design basis conditions. Treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations.

The distinction between treatment and special treatment is the degree of NRC specification as to what must be implemented for particular SSCs or for particular conditions. Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms: safety-related, important to safety, or basic component. The terms safety-related and basic component are defined in the regulations, while important to safety, used principally in the general design criteria (GDC) of Appendix A to 10 CFR Part 50, is not explicitly defined.

2.2 REASON FOR PROPOSED CHANGE A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, Probabilistic Risk Assessments (PRAs) address credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an

Enclosure to TXX-23016 Page 3 of 24 extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

To take advantage of the safety enhancements available through the use of PRA, in 2004 the NRC published a new regulation, 10 CFR 50.69. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with the regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The rule contains requirements on how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the SSC, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four risk-informed safety class (RISC) categories. The determination of safety significance is performed by an integrated decision-making process, as described by NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline (Reference 2), which uses both risk insights and traditional engineering insights. The safety functions include the design basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSCs is applied as necessary to maintain functionality and reliability and is a function of the SSC categorization results and associated bases. Finally, periodic assessment activities are conducted to make adjustments to the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable requirements.

The rule does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility. Instead, the rule enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as high safety significant, existing treatment requirements are maintained or enhanced. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows an alternative risk-informed approach to treatment that provides reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

Implementation of 10 CFR 50.69 will allow Vistra Operations Company LLC, (Vistra OpCo) to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE Vistra OpCo proposes the addition of the following condition to the operating licenses of Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (CPNPP), to document the NRCs approval of the use 10 CFR 50.69:

Vistra OpCo is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations

Enclosure to TXX-23016 Page 4 of 24 that are based on a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009.

In addition, Vistra OpCo is approved to implement 10 CFR 50.69 using the EPRI alternative Tier 1 seismic approach for active categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs, and as specified in License Amendment No. [XXX] dated [DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from the EPRI alternate seismic approach to a seismic probabilistic risk assessment approach).

3 TECHNICAL EVALUATION 10 CFR 50.69 specifies the information to be provided by a licensee requesting adoption of the regulation. This request conforms to the requirements of 10 CFR 50.69(b)(2), which states:

A licensee voluntarily choosing to implement this section shall submit an application for license amendment under § 50.90 that contains the following information:

(i) A description of the process for categorization of RISC-1, RISC-2, RISC-3 and RISC-4 SSCs.

(ii) A description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific probabilistic risk assessment (PRA), margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

(iii) Results of the PRA review process conducted to meet § 50.69(c)(1)(i).

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy

§ 50.69(c)(1)(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions).

Each of these submittal requirements are addressed in the following sections.

The NRC has previously reviewed the technical adequacy of the CPNPP PRA models identified in this application for internal events, including internal flooding, and internal fire in Reference 1.

Vistra OpCo requests that the NRC utilize the review of the PRA technical adequacy for that application when performing the review for this application.

3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(B)(2)(I))

3.1.1 Overall Categorization Process Vistra OpCo will implement the risk categorization process in accordance with NEI 00-04, Reference 2, as endorsed by Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety

Enclosure to TXX-23016 Page 5 of 24 Significance (Reference 3). NEI 00-04 Section 1.5 states, Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant. A separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The process to categorize each system will be consistent with the guidance in NEI 00-04, as endorsed by RG 1.201, with the exception of the evaluation of impact of the seismic hazard, which will use the EPRI Alternative Tier 1 Seismic Approach described in EPRI Report 3002022453 (Reference 5). RG 1.201 states that, the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(iv). However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence or order for each of the elements to be completed.

Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and as long as 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:

1. PRA-based evaluations (e.g., the internal events, internal flooding, and internal fire PRAs).
2. Non-PRA approaches (e.g., extreme winds and tornadoes, EPRI Alternative Seismic Approach, and other external events screening, and shutdown assessment).
3. Seven qualitative criteria in Section 9.2 of NEI 00-04.
4. The defense-in-depth assessment.
5. The passive categorization methodology.

Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various elements identified above. The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., High Safety Significant [HSS] or Low Safety Significant [LSS]) that is presented to the Integrated Decision-Making Panel (IDP). Note: the term preliminary HSS or LSS is synonymous with the NEI 00-04 term candidate HSS or LSS. A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1. The safety significance determination of each element, identified above, is independent of each other and therefore the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00-04, the categorization of a component or function will only be preliminary until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final Risk Informed Safety Class (RISC) category can be assigned.

The IDP may direct and approve detailed categorization of components in accordance with NEI 00-04 Section 10.2. The IDP may always elect to change a preliminary LSS component or function to HSS, however the ability to change component categorization from preliminary HSS to LSS is limited. This ability is only available to the IDP for select process steps as described in

Enclosure to TXX-23016 Page 6 of 24 NEI 00-04 and endorsed by RG 1.201. Table 3-1 summarizes these IDP limitations in NEI 00-04. The steps of the process are performed at either the function level, component level, or both. This is also summarized in Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

Table 3-1. Categorization Evaluation Summary Categorization Drives Element Step - NEI 00-04 Evaluation Level IDP Change HSS Associated Section to LSS Functions Internal Events Base Case - Not Allowed Yes Section 5.1 Fire, Seismic and other External Allowable No Risk (PRA Events Base Case Component Modeled)

PRA Sensitivity Allowable No Studies Integral PRA Assessment - Not Allowed Yes Section 5.6 Fire and Other Component Not Allowed No External Hazards Seismic -

Risk (Non-Alternative Tier 1 Function/Component Allowed(1) No modeled)

Approach Shutdown -

Function/Component Not Allowed No Section 5.5 Core Damage -

Function/Component Not Allowed Yes Defense-in- Section 6.1 Depth Containment -

Component Not Allowed Yes Section 6.2 Qualitative Considerations -

Function Allowable(2) N/A Criteria Section 9.2 Passive -

Passive Segment/Component Not Allowed No Section 4 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.

Enclosure to TXX-23016 Page 7 of 24 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 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.

The mapping of components to system functions is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., Internal events PRA or Integral PRA assessment) or defense-in-depth evaluation will be initially treated as HSS.

However, NEI 00-04, Section 10.2 allows detailed categorization which can result in some components mapped to HSS functions being treated as LSS; and Section 4.0 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with a HSS function but which do not support the critical attributes of that HSS function. Note that certain steps of the categorization process are performed at a component level (e.g., Passive, Non-PRA-modeled hazards - see Table 3-1). Except for seismic, these components from the component level assessments will remain HSS (IDP cannot override) regardless of the significance of the functions to which they are mapped. Components having seismic functions may be HSS or LSS based on the IDPs consideration of the seismic insights applicable to the system being categorized. Therefore, if a HSS component is mapped to an LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven HSS based on Table 3-1 or may remain LSS. For the seismic hazard, given that Comanche Peak is a seismic Tier 1 (low seismic hazard) plant as defined in EPRI 3002022453, seismic considerations are not required to drive an HSS determination at the component level, but the IDP will consider available seismic information pertinent to the components being categorized and can, at its discretion, determine that a component should be HSS based on that information.

NEI 00-04, Sections 4 and 7.1, will be followed for SSCs that support an interfacing system.

Those SSCs will typically remain uncategorized until all interfacing systems are categorized. In some cases, impacts that an interfacing component could have on an interfacing system can be fully determined and the interface component can be categorized (and alternative treatment implemented) without categorizing the entire interfacing system. In this event, an assessment of interface component risk associated with uncategorized systems will be limited to cases where the following two conditions are met: 1) the interface component failure cannot prevent performance of interface system functions, and 2) the risk is limited to passive failures assessed as LSS following the passive categorization process for the applicable pressure boundary segments.

Enclosure to TXX-23016 Page 8 of 24 The following are clarifications to be applied to the NEI 00-04 categorization process:

  • The Integrated Decision-Making Panel (IDP) will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) engineering, system engineering, safety analysis, and probabilistic risk assessment. At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA.
  • The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address at a minimum the purpose of the categorization; present treatment requirements for SSCs including requirements for design basis events; PRA fundamentals; details of the plant specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the defense-in- depth philosophy and requirements to maintain this philosophy.
  • The decision criteria for the IDP for categorizing SSCs as high safety significant or low safety significant pursuant to § 50.69(f)(1) will be documented in Vistra OpCo procedures.

Decisions of the IDP will be arrived at by consensus. Differing opinions will be documented and resolved, if possible. However, a simple majority of the panel is sufficient for final decisions regarding High Safety Significant (HSS) and Low Safety Significant (LSS).

  • Passive characterization will be performed using the processes described in subsection 3.1.2. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.
  • An unreliability factor of 3 will be used for the sensitivity studies described in Section 8 of NEI 00-04. The factor of 3 was chosen as it is representative of the typical error factor of basic events used in the PRA model.
  • NEI 00-04, Section 7 requires assigning the safety significance of functions to be preliminary HSS if it is supported by an SSC determined to be HSS from the PRA-based assessment in Section 5 but does not require this for SSCs determined to be HSS from non-PRA-based, deterministic assessments in Section 5. This requirement is further clarified in the Vogtle SER (Reference 8) which states if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6 of NEI 00-04), the associated system function(s) would be identified as HSS.
  • Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The IDP must intervene to assign any of these HSS Function components to LSS.
  • With regard to the criteria that considers whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, Vistra OpCo will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

Enclosure to TXX-23016 Page 9 of 24 The following are the exceptions taken to the NEI 00-04 categorization process:

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

  • Internal Event Risks: Internal events including internal flooding PRA model version accepted by NRC in subsection IEPRA in Section 3.2.4.1.2 of Reference 1.
  • Fire Risks: Fire PRA model versions accepted by NRC in subsection Internal Fire Events PRA in Section 3.2.4.1.2 of Reference 1.
  • Seismic Risks: EPRI Alternative Seismic Approach for Tier 1 plants identified in EPRI Report 3002022453 (Reference 5). Vistra OpCo will follow a similar alternative seismic approach in the 10 CFR 50.69 categorization process for CPNPP as the approach that was approved by the NRC staff for Calvert Cliffs Nuclear Power Plant (ADAMS Accession No. ML19330D909) (Reference 19).
  • Other External Risks (e.g., tornados, external floods):

- Extreme Wind and Tornado Hazards: Tornado Safe Shutdown Equipment List as discussed in Section 3.2.4 of this LAR.

- All other external hazards were determined to be insignificant contributors to plant risk as accepted by the NRC in Reference 1.

  • Low Power and Shutdown Risks: Qualitative defense-in-depth (DID) shutdown model for shutdown configuration risk management (CRM) based on the framework for DID provided in NUMARC 91-06, Guidance for Industry Actions to Assess Shutdown Management (Reference 6), which provides guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above (e.g., change from the EPRI alternate seismic approach to a seismic probabilistic risk assessment) will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization.
2. System functions, identified and categorized with the associated bases.
3. Mapping of components to support function(s).
4. PRA model results, including sensitivity studies.
5. Hazards analyses, as applicable.
6. Passive categorization results and bases.
7. Categorization results including all associated bases and RISC classifications.

Enclosure to TXX-23016 Page 10 of 24

8. Component critical attributes for HSS SSCs.
9. Results of periodic reviews and SSC performance evaluations.
10. IDP meeting minutes and qualification/training records for the IDP members.

3.1.2 Passive Categorization Process For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Arkansas Nuclear One (ANO) Risk-Informed Repair/Replacement Activities (RI-RRA) methodology contained in Reference 7, consistent with the related Safety Evaluation (SE) issued by the Office of Nuclear Reactor Regulation.

The RI-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities (RI-RRA methodology) for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., DID, safety margins) in determining safety significance. Component supports, if categorized, are assigned the same safety significance as the highest passively ranked component within the bounds of the associated analytical pipe stress model. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

The use of this method was previously approved to be used for a 10 CFR 50.69 application by NRC in the final Safety Evaluation for Vogtle (Reference 8). The RI-RRA method as approved for use at Vogtle for 10 CFR 50.69 does not have any plant specific aspects and is generic. It relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to changes in treatment. The passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the ANO2-R&R-004 for the passive categorization of Class 2, 3, and non-class components. This is the same passive SSC scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in Regulatory Guide 1.147, Revision 15 (Reference 9). Both code cases employ a similar risk-informed safety classification of SSCs in order to change the repair/

replacement requirements of the affected LSS components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned high safety-significant, HSS, for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP. Therefore, this methodology and scope for passive categorization is acceptable and appropriate for use at CPNPP for 10 CFR 50.69 SSC categorization.

3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(B)(2)(II))

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. The PRA models credited in this request are the same PRA models credited in the TSTF-505, Revision 2 application dated May 11, 2021 for internal

Enclosure to TXX-23016 Page 11 of 24 events, internal floods and internal fire (Reference 10) accepted by the NRC in Reference 1.

3.2.1 Internal Events and Internal Flooding The CPNPP categorization process for the internal events and internal flooding hazard will use the plant-specific PRA model. The Vistra OpCo risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the Comanche Peak units. Attachment 2 of this enclosure identifies the results of the applicable internal events and internal flooding PRA models.

3.2.2 Fire Hazards The CPNPP categorization process for fire hazards will use a peer reviewed plant-specific fire PRA model. The internal fire PRA model has been developed consistent with NUREG/CR-6850 (Reference 4) and only utilizes methods previously accepted by the NRC. The Vistra OpCo risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the Comanche Peak units. Attachment 2 of this enclosure identifies the results of the applicable fire PRA model.

3.2.3 Seismic Hazards 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 Reference 5, where CPNPP meets the 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 3002022453 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).

This approach 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 Reference 5 along with confirmation that the site GMRS is low. Reference 5 demonstrates that seismic risk is adequately addressed for Tier 1 sites by the results of the other elements of the 50.69 categorization process.

As an 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

Enclosure to TXX-23016 Page 12 of 24 condition of HSS if the results of the integral assessment 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.

Reference 5 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 Reference 5 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 Reference 5:

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.
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

Enclosure to TXX-23016 Page 13 of 24 the equipment is HSS only due to seismic considerations.

Test cases described in Section 3 of Reference 5 showed that it would be unusual even for moderate hazard plants to exhibit any unique seismic insights, including due to correlated failures. 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 3002022453 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 assumed 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 following provides the basis for concluding that CPNPP meets the Tier 1 site criteria:

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

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 Engineering, 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

Enclosure to TXX-23016 Page 14 of 24 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.

3.2.4 Other External Hazards The Other External Hazards credited in this request addressed below are the same non-PRA hazards in the TSTF-505, Revision 2 application (Reference 10), and approved in Reference 1, Section 3.2.4.1.3, subsections Evaluation of Extreme Winds and Tornado Hazards and Evaluation of Other External Hazards.

Extreme Winds and Tornados Both extreme winds and tornados are examined in the tornado hazard evaluation. Since the tornado hazard is not screened, CPNPP categorization process will use the process described below to determine the safety significance of SSCs for the tornado hazard. The hazard is assumed to be present during a tornado-induced loss of offsite power.

The tornado hazard safety significance process uses a Tornado Safe Shutdown Equipment List (TSSEL) of SSCs that was developed from a list of SSCs needed to achieve and maintain safe shutdown of the reactor assuming unavailability of offsite power. During categorization of systems, the NEI 00-04 component to function mapping process will be applied to the safe shutdown function of (1) Decay Heat Removal, (2) Reactivity Control, (3) Inventory Control, (4)

Power Availability, and (5) Reactor Pressure Control. The SSCs that fulfill the tornado safe shutdown functions, as well as any tornado barriers that are credited with protecting equipment that fulfills a TSSEL function, will be identified as candidate HSS for the system being categorized regardless of their tornado damage susceptibility or frequency of challenge. This approach ensures the SSCs that are credited to achieve and maintain the capability for safe shutdown are retained as safety significant.

The safety significance process for the tornado hazard is shown in Figure 3-1. There are no importance measures used in determining safety significance of SSCs related to the tornado hazard. As stated in NEI 00-04, an SSC identified as HSS by a non-PRA method for external events "may not be re-categorized by the IDP."

Enclosure to TXX-23016 Page 15 of 24 Figure 3-1 Safety Significance Process for Systems and Components for the Tornado Protection Program Select SSC No No Does the SSC support a TSSEL Candidate Low Safety Is the SSC on the TSSEL?

Function? Significant Yes Candidate High Safety Significant Identify Safety Significant Attributes of Component All Other External Hazards All remaining hazards were screened from applicability and considered insignificant for every SSC and, therefore , will not be considered during the categorization process. Other external hazards were evaluated in TSTF-505 , Revision 2 and were screened as identified in Reference 1, Section 3.2.4.1.3, subsection Evaluation of Other External Hazards.

3.2.5 Low Power & Shutdown Consistent with NEI 00-04, the CPNPP categorization process will use the shutdown safety management plan described in NUMARC 91-06 for evaluation of safety significance related to low power and shutdown conditions. The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04.

NUMARC 91-06 specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function . The key safety functions defined in NU MARC 91-06 are evaluated for categorization of SSCs.

SSCs that meet either of the two criteria (i .e., considered part of a "primary shutdown safety system" or a failure would initiate an event during shutdown conditions) described in Section 5.5 of NEI 00-04 will be considered preliminary HSS.

Enclosure to TXX-23016 Page 16 of 24 3.2.6 PRA Maintenance and Updates The Vistra OpCo risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant for each of the Comanche Peak units. The process delineates the responsibilities and guidelines for updating the PRA models and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, and industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The process will assess the impact of these changes on the plant PRA model in a timely manner but no longer than once every 48 months consistent with the approved TSTF-505, Revision 2 identified in Section 3.2.4.1.7 in Reference 1. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

In addition, Vistra OpCo will implement a process that addresses the requirements in NEI 00-04, Section 11, Program Documentation and Change Control. The process will review the results of periodic and interim updates of the plant PRA that may affect the results of the categorization process. If the results are affected, adjustments will be made as necessary to the categorization or treatment processes to maintain the validity of the processes. In addition, any PRA model upgrades will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

3.2.7 PRA Uncertainty Evaluations Uncertainty evaluations associated with any applicable baseline PRA model(s) used in this application were evaluated during the assessment of PRA technical adequacy and confirmed through the self-assessment and peer review processes as accepted by the NRC in the approval of TSTF-505, Revision 2 identified in Section 3.2.4.1.5 in Reference 1.

Uncertainty evaluations associated with the risk categorization process are addressed using the processes discussed in Section 8 of NEI 00-04 and in the prescribed sensitivity studies discussed in Section 5 of NEI 00-04.

In the overall risk sensitivity studies, Vistra OpCo will use a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference 8. Consistent with the NEI 00-04 guidance, Vistra OpCo will perform both an initial sensitivity study and a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in all identified PRA models for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

The detailed process of identifying, characterizing and qualitative screening of model uncertainties is found in Section 5.3 of NUREG-1855 (Reference 12) and subsection 3.1.1 of EPRI TR-1016737 (Reference 13). The process in these references was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups.

Enclosure to TXX-23016 Page 17 of 24 Each PRA element notebook has been reviewed for assumptions and sources of uncertainties.

The characterization of assumptions and sources of uncertainties are based on whether the assumption and/or source of uncertainty is key to the 50.69 application in accordance with RG 1.200 Revision 2 (Reference 11).

Key CPNPP PRA model specific assumptions and sources of uncertainty were identified and dispositioned for the RICT calculation as identified in Section 3.2.4.1.5 in Reference 1. The conclusion of this review is that no additional sensitivity analyses are required to address CPNPP PRA model specific assumptions or sources of uncertainty.

3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(B)(2)(III))

The PRA models described in Section 3.2 has been assessed against RG 1.200, (Reference 11) consistent with NRC RIS 2007-06 (Reference 14).

Section 3.2.4.1.1 in the NRC approval of TSTF-505, Revision 2 (Reference 1) demonstrates that the PRA is of sufficient quality and level of detail to support risk informed applications and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required 10 CFR 50.69(c)(1)(i).

3.4 RISK EVALUATIONS (10 CFR 50.69(B)(2)(IV))

The CPNPP 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04, with exception for the seismic hazard which will implement the EPRI Alternative Tier 1 Seismic Approach described in EPRI Report 3002022453. The overall risk evaluation process described in the NEI 00-04 guidance addresses both known degradation mechanisms and common cause interactions and meets the requirements of §50.69(b)(2)(iv).

Sensitivity studies described in NEI 00-04, Section 8 will be used to confirm that the categorization process results in acceptably small increases to CDF and large early release frequency (LERF). The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors). Subsequent performance monitoring and PRA updates required by the rule will continue to capture this data and provide timely insights into the need to account for any important new degradation mechanisms.

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 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 (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,

Enclosure to TXX-23016 Page 18 of 24 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
  • A review of the importance measures used for screening in the categorization process
  • An update of the risk sensitivity study performed for the categorization In addition to the normally scheduled periodic reviews, if a PRA model or other risk information is upgraded, a review of the SSC categorization will be performed.

4 REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The following NRC requirements and guidance documents are applicable to the proposed change.

  • The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors.
  • 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. (Reference 3)

(Reference 11)

The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS Vistra OpCo proposes to modify the CPNPP licensing basis to allow for the voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety

Enclosure to TXX-23016 Page 19 of 24 significance resulting in improved plant safety.

Vistra OpCo has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special

Enclosure to TXX-23016 Page 20 of 24 treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Vistra OpCo concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.3 CONCLUSION

S In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6 REFERENCES

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.

Enclosure to TXX-23016 Page 21 of 24

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/Replacement 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 Operating 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 Inspection Code Case Acceptability, ASME Section XI, Division 1, October 2007.
10. Letter from Vistra OpCo to the NRC, Application 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.
14. NRC Regulatory Issue Summary 2007-06, Regulatory 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 Applications, 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.

Enclosure to TXX-23016 Page 22 of 24

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).

Enclosure to TXX-23016 Page 23 of 24 Attachment 1: List of Categorization Prerequisites Vistra OpCo will establish procedure(s) prior to the use of the categorization process on a plant system. The procedure(s) will contain the elements/steps listed below:

  • Integrated Decision-Making Panel (IDP) member qualification requirements
  • Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary High Safety Significant (HSS) or Low Safety Significant (LSS) based on the seven criteria in Section 9 of NEI 00-04 (see Section 3.1). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting an LSS function are categorized as preliminary LSS.
  • Component safety significance assessment. Safety significance of active components is assessed through a combination of Probabilistic Risk Assessment (PRA) and non-PRA methods, covering all hazards. Safety significance of passive components is assessed using a methodology for passive components.
  • Assessment of defense-in-depth (DID) and safety margin. Safety-related components that are categorized* as preliminary LSS are evaluated for their role in providing DID and safety margin and, if appropriate, upgraded to HSS.
  • Review by the IDP. The categorization results are presented to the lDP for review and approval. The lDP reviews the categorization results and makes the final determination on the safety significance of system functions and components.
  • Risk sensitivity study. For PRA-modeled components, an overall risk sensitivity study is used to confirm that the population of preliminary LSS components results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF) and meets the acceptance guidelines of Regulatory Guide 1.174.
  • Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized.
  • Documentation requirements per subsection 3.1.1 of the enclosure.

Enclosure to TXX-23016 Page 24 of 24 Attachment 2: CPNPP Baseline PRA Model/Non-PRA Results CPNPP Baseline PRA Model Results(1) (per reactor year of operation)

CPNPP Unit 1 Baseline CDF Baseline LERF Internal Events 1.10E-06 Internal Events 1.06E-07 Internal Flood 1.19E-07 Internal Flood 5.00E-09 Internal Fire 5.62E-05 Internal Fire 7.89E-06 Seismic(2) 1.93E-06 Seismic(2) 9.73E-07 High Winds(3) 4.09E-06 High Winds(3) 2.35E-07 No significant No significant Other External Events Other External Events contribution contribution Total Unit 1 CDF 6.34E-05 Total Unit 1 LERF 9.21E-06 CPNPP Unit 2 Baseline CDF Baseline LERF Internal Events 1.02E-06 Internal Events 1.02E-07 Internal Flood 1.39E-07 Internal Flood 5.88E-09 Internal Fire 4.29E-05 Internal Fire 5.69E-06 Seismic(2) 1.93E-06 Seismic(2) 9.73E-07 High Winds (3) 4.09E-06 High Winds (3) 2.35E-07 No significant No significant Other External Events Other External Events contribution contribution Total Unit 2 CDF 4.98E-05 Total Unit 2 LERF 6.99E-06 Notes:

(1) These are the CPNPP Baseline PRA Model Results provided for RICT in the CPNPP APLC RAI Response (Reference 17). In the TSTF-505, Revision 2 NRC approval in Section 3.2.4.1.4 in Reference 1, it was stated that the current total CDF and LERF estimated using the Comanche Peak PRAs and including the penalty factors for seismic and high winds risk, and after accounting for SOKC in the quantification of the PRAs, meet the RG 1.174, Revision 3 guidelines.

(2) The baseline seismic CDF and LERF are based on the penalty factor used in the RICT calculation.

This value includes a unit power factor of 0.94. A seismic PRA is not used in the 10 CFR 50.69 categorization process and is instead represented here to show the current baseline CDF and LERF compliance with RG 1.174.

(3) The baseline high winds CDF and LERF are developed and are representative of a penalty factor used in the RICT calculation. This value includes a unit power factor of 0.94. The high winds penalty is procedurally adjusted based on the component(s) to which the RICT is applied. A high winds PRA is not used in the 10 CFR 50.69 categorization process and is instead represented here to show the current baseline CDF and LERF compliance with RG 1.174.

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