Text

Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors
	ML23048A022
Person / Time
Site:	Mcguire, McGuire
Issue date:	02/17/2023
From:	Pigott E; Duke Energy Carolinas
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	RA-18-0090
	Download: ML23048A022 (1)
	v • d • e

fa DUKE Edward R. Pigott Site Vice President

~ ENERGY McGuire Nuclear Station Duke Energy MG01VP l 12700 Hagers Ferry Road Huntersville, NC 28078 o: 980.875.4111 Edward.Pigott@duke-energy.com February 17, 2023 Serial: RA-18-0090 10 CFR 50.90 10 CFR 50.69 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370, Renewed License Nos. NPF-9 and NPF-17

Subject:

Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors Ladies and Gentlemen:

In accordance with the provisions of Sections 50.90 and 50.69 of Title 10 of the Code of Federal Regulations (10 CFR), Duke Energy Carolinas, LLC (Duke Energy) is requesting an amendment to the renewed facility operating licenses (FOL) of McGuire Nuclear Station (MNS), Units 1 and 2.

The proposed amendment would modify the MNS licensing basis, by the addition of a License Condition, to allow for the implementation of the provisions of 10 CFR 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 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 change to the MNS, Units 1 and 2 FOLs. The categorization process being implemented through this change is consistent with Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, dated July 2005, which was endorsed by the 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. 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.

The probabilistic risk assessment (PRA) models described within this license amendment request (LAR) are the same as those described within the Duke Energy submittal of the MNS LAR dated February 16, 2023 for the adoption of Technical Specifications Task Force (TSTF)

Traveler 505 (Serial: RA-18-0190). Duke Energy requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the

U.S. Nuclear Regulatory Commission Page2 TSTF-505 application that is also currently under review. This would reduce the number of Duke Energy and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

Duke Energy requests approval of the proposed license amendment 12 months following acceptance, with an implementation period of 120 days.

There are no regulatory commitments made in this submittal.

In accordance with 10 CFR 50.91 (a)(1 ), "Notice for Public Comment," the analysis about the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is being provided to the Commission .

In accordance with 10 CFR 50.91(b)(1), "Notice for Public Comment; State Consultation," a copy of this application, with the enclosure and attachments, is being provided to the designated North Carolina Official.

Please refer any questions regarding this submittal to Mr. Ryan Treadway, Director, Nuclear Fleet Licensing, at (980) 373-5873.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on February 17, 2023.

Edward R. Pigott Site Vice President McGuire Nuclear Station

Enclosure:

Description and Assessment of the Proposed Change Attachments:

1. List of Categorization Prerequisites

2. Description of PRA Models Used in Categorization

3. Disposition and Resolution of Open Peer Review Findings and Self-Assessm ent Open Items

4. External Hazards Screening

5. Progressive Screening Approach for Addressing External Hazards

6. Disposition of Key Assumptions/S ources of Uncertainty

7. Markup of McGuire, Units 1 and 2 Renewed Facility Operating Licenses

U.S. Nuclear Regulatory Commission Page 3 cc (w/ Enclosure and Attachments):

L. Dudes, Regional Administrator, Region II J. Klos, NRR Project Manager C. Safouri, NRC Senior Resident Inspector D. Crowley, Interim Section Chief, Radiation Protection Section, NC DHHS

U.S. Nuclear Regulatory Commission Page 1 RA-18-0090 Enclosure Description and Assessment of the Proposed Change

Subject:

Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 Current Regulatory Requirements 2.2 Reason for the Proposed Change 2.3 Description of the Proposed Change

3. TECHNICAL EVALUATION 3.1 Categorization Process Description (10 CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process 3.1.2 Passive Categorization Process 3.2 Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii))

3.2.1 Internal Events and Internal Flooding 3.2.2 Fire Hazards 3.2.3 Seismic Hazards 3.2.4 Other External Hazards 3.2.5 Low Power & Shutdown 3.2.6 PRA Maintenance and Updates 3.2.7 PRA Uncertainty Evaluations 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 Determination Analysis 4.3 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES ATTACHMENTS:

1. List of Categorization Prerequisites

2. Description of PRA Models Used in Categorization

3. Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items

4. External Hazards Screening

5. Progressive Screening Approach for Addressing External Hazards

6. Disposition of Key Assumptions/Sources of Uncertainty

7. Markup of McGuire, Units 1 and 2 Renewed Facility Operating Licenses

U.S. Nuclear Regulatory Commission Page 2 RA-18-0090

1.

SUMMARY

DESCRIPTION The proposed amendment modifies the licensing basis 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 the 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

U.S. Nuclear Regulatory Commission Page 3 RA-18-0090 an 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 Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline (Reference 1), 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 Duke Energy Carolinas, LLC (Duke Energy) to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 Description of the Proposed Change Duke Energy proposes the addition of the following condition to the renewed facility operating licenses (FOL) of MNS Units 1 and 2 to document the NRC's approval of the use of 10 CFR 50.69.

Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC3, 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

U.S. Nuclear Regulatory Commission Page 4 RA-18-0090 component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic, and the alternative seismic approach described in Duke Energys submittal letter RA-18-0090 dated February 17, 2023; 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 a seismic margins approach to a seismic probabilistic risk assessment approach).

A markup of the MNS, Units 1 and 2 FOLs to reflect the proposed change is provided in .

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 PRA models described within this license amendment request (LAR) are the same as those described within the Duke Energy submittal of the MNS LAR dated February 16, 2023 for the adoption of Technical Specifications Task Force (TSTF) Traveler 505 (Serial: RA-18-0190).

Duke Energy requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the application currently in-process.

This would reduce the number of Duke Energy and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR are complete which will allow the

U.S. Nuclear Regulatory Commission Page 5 RA-18-0090 NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

3.1 Categorization Process Description (10 CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process Duke Energy will implement the risk categorization process in accordance with NEI 00-04, Revision 0, as endorsed by Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance (Reference 2). 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, 10 CFR 50.69 SSC Categorization Guideline, as endorsed by RG 1.201, with the exception of:

x The evaluation of impact of the seismic hazard, which will use the Electric Power Research Institute (EPRI) 3002017583 (Reference 17) approach for seismic Tier 2 sites, which includes MNS, to assess seismic hazard risk for 10 CFR 50.69.

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 Low Safety Significant (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. Inclusion of additional process steps discussed below to address seismic considerations will ensure that reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(iv) is achieved.

1. PRA-based evaluations (e.g., the internal events, internal flooding, and fire PRAs)

2. Non-PRA approaches (e.g., 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 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

U.S. Nuclear Regulatory Commission Page 6 RA-18-0090 determination in accordance with Table 3-1 below. 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 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 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 the Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

Table 3-1: Categorization Evaluation Summary Drives Categorization Step IDP Change Element Evaluation Level Associated

- NEI 00-04 Section HSS 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 Other External Component Not Allowed No Hazards Risk (Non- Seismic -

modeled) Alternative Tier 2 Function/Component Allowed1 No 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 Allowable2 N/A Criteria Section 9.2 Passive Passive - Section 4 Segment/Component Not Allowed No

U.S. Nuclear Regulatory Commission Page 7 RA-18-0090 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 Section 9.2. In some cases, a 10 CFR 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 10 CFR 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 10 CFR 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 10 CFR 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 above or may remain LSS. For the seismic hazard, given that MNS is a seismic Tier 2 (moderate seismic hazard) plant as defined in Reference 17, 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.

The following are clarifications to be applied to the NEI 00-04 categorization process:

x The 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 PRA. 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

U.S. Nuclear Regulatory Commission Page 8 RA-18-0090 minimum of three years of experience in the modeling and updating of the plant-specific PRA.

x 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.

x The decision criteria for the IDP for categorizing SSCs as safety significant or low safety-significant pursuant to § 50.69(f)(1) will be documented in Duke Energy procedures.

x Decisions of the IDP will be arrived at by consensus.

x Differing opinions will be documented and resolved, if possible. However, a simple majority of the panel is sufficient for final decisions regarding HSS and LSS.

x Passive characterization will be performed using the process described in Section 3.1.2.

Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

x 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.

x 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 Safety Evaluation (SE) (Reference 4) 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), the associated system function(s) would be identified as HSS.

x 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.

x With regard to the criteria that considers whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, Duke Energy will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

x Duke Energy proposes to apply an alternative seismic approach to those listed in NEI 00-04 Sections 1.5 and 5.3, as discussed in Section 3.2.3 of this LAR.

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

U.S. Nuclear Regulatory Commission Page 9 RA-18-0090 x Internal Event Risks: Internal events including internal flooding PRA, as submitted to the NRC for the adoption of TSTF 505 by letter dated February 16, 2023 (Refer to Attachment 2).

x Fire Risks: Fire PRA model, as submitted to the NRC for the adoption of TSTF 505 by letter dated February 16, 2023 (Refer to Attachment 2).

x Seismic Risks: EPRI Alternative Approach in EPRI 3002022453 (Reference 17) for Tier 2 plants and additional considerations discussed in Section 3.2.3 of this LAR.

x Other External Risks (e.g., tornados, external floods):

x Using the IPEEE screening process as approved by NRC SE dated February 16, 1999 (Reference 16). The other external hazards were determined to be insignificant contributors to plant risk.

x 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 3), 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 a seismic margins approach to a seismic PRA approach) 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

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 5 consistent with the related 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

U.S. Nuclear Regulatory Commission Page 10 RA-18-0090 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 SE for Vogtle dated December 17, 2014 (Reference 4). 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 can apply the same risk-informed process accepted by the NRC in Reference 5 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, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 15. 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 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 MNS 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 described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed. The PRA models credited in this request are the same PRA models credited in the MNS application to adopt TSTF-505 dated February 16, 2023.

3.2.1 Internal Events and Internal Flooding The MNS categorization process for the internal events and flooding hazard will use the plant-specific PRA model. The Duke Energy risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the MNS units. of this enclosure identifies the applicable internal events and internal flooding PRA models.

3.2.2 Fire Hazards The MNS categorization process for fire hazards will use a peer reviewed plant-specific fire PRA model. The internal Fire PRA model was developed consistent with NUREG/CR-6850 and only utilizes methods previously accepted by the NRC. The Duke Energy risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the MNS units. Attachment 2 of this enclosure identifies the applicable Fire PRA model.

U.S. Nuclear Regulatory Commission Page 11 RA-18-0090 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 (Reference 6) 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 MNS seismic hazard assessment, Duke Energy 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 17 with the EPRI markups provided in Attachment 2 of References 18 and 19 and includes additional considerations that are discussed in this section.

Note: The discussion below pertaining to Reference 17 includes the markups provided in Attachment 2 of References 18 and 19.

EPRI 3002017583 (Reference 17) is an update to EPRI 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," July 2018 (Reference 57) which was referenced in the NRC-issued amendment and SE for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, to implement 10 CFR 50.69 as noted below:

(1) 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 (Reference 58).

(2) This license amendment incorporated by reference the Clinton Power Station, Unit 1 response to request for additional information DRA/APLC RAI 03 - Alternate Seismic Approach included in the letter dated November 24, 2020 (Reference 59), in particular, the response to the question regarding the differences between the initial EPRI report 3002012988 and the current EPRI report 3002017583.

The proposed categorization approach for MNS is a risk-informed graded approach that is demonstrated to produce categorization insights equivalent to a seismic PRA. This approach relies on the insights gained from the seismic PRAs examined in Reference 17 and plant specific insights considering seismic correlation effects and seismic interactions. Following the criteria in Reference 17, the MNS site is considered a Tier 2 site because the site Ground Motion Response Spectrum (GMRS) to Safe Shutdown Earthquake (SSE) comparison is above the Tier 1 threshold but not high enough that the NRC required the plant to perform a seismic PRA (SPRA) to respond to Recommendation 2.1 of the Near Term Task Force 50.54(f) letter (Reference 20). Reference 17 also demonstrates that seismic risk is adequately addressed for Tier 2 sites by the results of additional qualitative assessments discussed in this section and existing elements of the 10 CFR 50.69 categorization process specified in NEI 00-04.

The trial studies in Reference 17, as amended by their request for additional information (RAI) responses and amendments (References 21, 22, 23, 24, 25, 26, 27, 28, and 29) 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 2 classification and resulting criteria is that consideration of the full range of the seismic hazard produces

U.S. Nuclear Regulatory Commission Page 12 RA-18-0090 limited unique insights to the categorization process. That is the basis for the following statements in Table 4-1 of Reference 17.

"At Tier 2 sites, there may be a limited number of unique seismic insights, most likely attributed to the possibility of seismically correlated failures, appropriate for consideration in determining HSS SSCs. The special seismic risk evaluation process recommended using a Common Cause impact approach in the FPIE [full power internal events] PRA can identify the appropriate seismic insights to be considered with the other categorization insights by the Integrated Decision-making Panel for the final HSS determinations."

At sites with moderate seismic demands (i.e., Tier 2 range) such as MNS, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference 30. Tier 2 seismic demand sites have a lower likelihood of seismically induced failures and less challenges to plant systems than trial study plants.

This, therefore, provides the technical basis for allowing use of a graded approach for addressing seismic hazards at MNS.

Test cases described in Section 3 of Reference 17, as amended by their RAI responses and amendments (References 21, 22, 23, 24, 25, 26, 27, 28, and 29) showed that there are very few, if any, SSCs that would be designated HSS for seismic unique reasons. The test cases identified that the unique seismic insights were typically associated with seismically correlated failures and led to unique HSS SSCs. While it would be unusual even for moderate hazard plants to exhibit any unique seismic insights, it is prudent and recommended by Reference 17 to perform additional evaluations to identify the conditions where correlated failures and seismic interactions may occur and determine their impact in the 10 CFR 50.69 categorization process. The special sensitivity study recommended in Reference 17 uses common cause failures, similar to the approach taken in a FPIE PRA and can identify the appropriate seismic insights to be considered with the other categorization insights by the IDP for the final HSS determinations.

Duke Energy is using test case information from Reference 17, developed by other licensees.

The test case information is being incorporated by reference into this application, specifically Case Study A (Reference 31), Case Study C (Reference 32), and Case Study D (Reference 33), as well as RAI responses and amendments (References 21, 22, 23, 24, 25, 26, 27, 28, and 29) that clarify aspects of these case studies.

Basis for MNS being a Tier 2 Plant As defined in Reference 17, MNS meets the Tier 2 criteria for a "Moderate Seismic Hazard /

Moderate Seismic Margin" site. The Tier 2 criteria are as follows:

"Tier 2: Plants where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is greater than in Tier 1 but not high enough to be treated as Tier 3. At these sites, the unique seismic categorization insights are expected to be limited."

Note: Reference 17 applies to the Tier 2 sites in its entirety except for Sections 2.2 (Tier 1 sites) and 2.4 (Tier 3 sites).

For comparison, Tier 1 plants are defined as having a GMRS peak acceleration at or below approximately 0.2g or where the GMRS is below or approximately equal to the SSE between

U.S. Nuclear Regulatory Commission Page 13 RA-18-0090 1.0 Hz and 10 Hz. Tier 3 plants are defined where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is high enough that the NRC required the plant to perform an SPRA to respond to the Fukushima 10 CFR 50.54(f) letter (Reference 20).

The NRC did not require MNS to perform an SPRA as stated in its revised seismic screening and prioritization letter dated December 22, 2016 (Reference 37):

"the NRC has determined that seismic probabilistic risk assessments (SPRAs) for Catawba and McGuire are no longer necessary to fulfill the March 12, 2012, request for information pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 50.54(f) (ADAMS Accession No. ML12053A340)."

The letter further states:

"the staff concludes that the plant-specific combination of seismic hazard exceedances, the general estimation of the seismic core damage frequencies for Catawba and McGuire, and insights related to the conditional containment failure probabilities at both these plants indicate that the increase in seismic risk due to the reevaluated seismic hazard is adequately addressed within the margin inherent in the design of these plants and, as such, the completion of SPRAs is not necessary."

As shown in Figure 1, comparing the MNS GMRS (derived from the seismic hazard) to the SSE (seismic design basis capability), the GMRS exceeds the SSE above 6 Hz. As such, it is appropriate that MNS is considered a Tier 2 plant. The basis for MNS being classified as Tier 2 will be documented and presented to the MNS IDP for each system that is categorized.

McGuire Response Spectra 1.00 0 .95 0 .90 0.85 0 .80 0 .75

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Figure 1: GMRS and SSE Response Spectra for McGuire (From Reference 35)

The following paragraphs describe additional background and the process to be utilized for the graded approach to categorize the seismic hazard for a Tier 2 plant.

U.S. Nuclear Regulatory Commission Page 14 RA-18-0090 Implementation of the Recommended Process Reference 17 recommends a risk-informed graded approach for addressing the seismic hazard in the 10 CFR 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 report. The coupling of these concepts with the categorization process in NEI 00-04 are the key elements of the approach defined in Reference 17 for identifying unique seismic insights.

The seismic fragility of an SSC is a function of the margin between an SSC's seismic capacity and the site-specific seismic demand. References such as EPRI NP-6041 (Reference 30) 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.

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.

In applying the Reference 17 process for Tier 2 sites to the MNS 10 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the Reference 17 guidance and informed of plant SSC-specific seismic insights that the IDP may choose to consider in their HSS/LSS deliberations. As part of the categorization team's preparation of the System Categorization document (SCD) that is presented to the IDP, a section will be included that provides identified plant seismic insights as well as the basis for applicability of the Reference 17 study and the bases for MNS being a Tier 2 plant. The discussion of the Tier 2 bases will include such factors as:

The moderate seismic hazard for the plant,

The definition of Tier 2 in the EPRI study, and

The basis for concluding MNS is a Tier 2 plant.

At several steps of the categorization process the categorization team will consider the available seismic insights relative 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 MNS) 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 MNS 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

U.S. Nuclear Regulatory Commission Page 15 RA-18-0090 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 these for presentation to the IDP as additional qualitative inputs, which will also be described in the SCD.

The categorization team will review available MNS plant-specific seismic reviews and other resources such as those identified above. The objective of the seismic review is to identify plant-specific seismic insights that might include potentially important impacts such as:

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

For each system categorized, the categorization team will evaluate correlated seismic failures and seismic interactions between SSCs. This process is detailed in Section 2.3.1 of Reference 17 and is summarized below in Figure 2.

U.S. Nuclear Regulatory Commission Page 16 RA-18-0090 2

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Figure 2: Seismic Correlated Failure Assessment for Tier 2 Plants (Reproduced from Reference 17: Figure 2-4)

Determination of seismic insights will make use of the full power internal events PRA model supplemented by focused seismic walkdowns. An overview of the process to determine the importance of SSCs for mitigating seismic events follows and is utilized on a system basis:

x Gather the population of SSCs in the system being categorized and review existing seismic information (reference Step 1 of Figure 2). This step may use the results of the

U.S. Nuclear Regulatory Commission Page 17 RA-18-0090 required Tier 1 assessment that is performed along with the Tier 2 assessment. As stated in Reference 17 the technical basis for the Tier 1 approach in Section 2.2 of Reference 17 generally applies for Tier 2 plants in addition to the additional sensitivity and walkdowns described herein.

x Assign seismic based SSC equipment class and distributed system IDs, as used for SPRAs, for SSCs in the system being categorized (reference Step 2 of Figure 2).

x Perform a series of screenings to refine the list of SSCs subject to correlation sensitivity studies. Screens will identify (reference Steps 3a/3b/3c of Figure 2):

o Inherently rugged SSCs o SSCs not in Level 1 or Level 2 PRAs o Components already identified as HSS components from the Internal Events PRA or Integrated assessment o The above screened SSCs will still be evaluated for seismic interactions.

x SSCs identified in the above screening can be screened from consideration as functional correlation surrogate events. They are removed from the remainder of the process (can be considered LSS) unless they are subject to interaction source considerations (reference Step 4 of Figure 2).

x Perform Tier 2 Walkdown(s) focusing on identifying seismic correlated or interaction SSC failures (reference Steps 5a/5b of Figure 2).

x Screen out from further seismic considerations SSCs that are determined through the walkdown to be of high seismic capacity and not included in seismically correlated groups or correlated interaction groups since their non-seismic failure modes are already addressed for 50.69 categorization in the FPIE PRA and Fire PRA. Those remaining components proceed forward for inclusion of associated seismic surrogate events in the Tier 2 Adjusted PRA Model (reference Steps 5c/6 of Figure 2).

x Develop a Tier 2 Adjusted PRA Model and incorporate seismic surrogate events into the model to reflect the potential seismically correlated and interaction conditions identified in prior steps (reference Steps 6/7 of Figure 2). The seismic surrogate basic events shall be added to the PRA under the appropriate areas in the logic model (e.g.,

given that the Tier 2 Adjusted PRA Model uses only loss of offsite power (LOOP) and small loss of coolant accident (LOCA) sequences, the seismic surrogate events should be added to system and/or nodal fault tree structures that tie into these sequence types). The probability of each seismic surrogate basic event added to the model should be set to 1.0E-04 (based on guidance in Reference 17).

x Quantify only the LOOP and small LOCA initiated accident sequences of the Tier 2 Adjusted PRA Model (reference Step 8 of Figure 2). The event frequency of the LOOP initiator shall be set to a value of 1.0 and the event frequency for the small LOCA initiator shall be set to a value of 1.0E-02. Remove credits for restoration of offsite power and other functional recoveries (e.g., Emergency Diesel Generator (EDG) and DC power recovery).

U.S. Nuclear Regulatory Commission Page 18 RA-18-0090 x Utilize the Importance Measures from the quantification of the Tier 2 Adjusted PRA Model to identify appropriate SSCs (in the system being categorized) that should be HSS due to correlation or seismic interactions (reference Step 9 of Figure 2).

x SSCs screened out in Steps 5c, 6, or 9 in Figure 2 above can be considered LSS (reference Step 10 of Figure 2).

x Prepare documentation of the Tier 2 analysis results, including identification of seismic unique HSS SSCs, for presentation to the IDP (reference Step 11 of Figure 2).

Seismic impacts would be compiled on an SSC basis. As each system is categorized, the system-specific seismic insights will be documented in the categorization report and provided to the IDP for consideration as part of the IDP review process. The IDP cannot challenge any candidate HSS recommendation for any SSC from a seismic perspective if they believe there is a basis, except for certain conditions identified in Step 10 of Section 2.3.1 of Reference 17.

Any decision by the IDP to downgrade preliminary HSS components to LSS will consider the applicable seismic insights in that decision. SSCs identified from the Fire PRA as candidate HSS, which are not HSS from the internal events PRA or integrated importance measure assessment, will be reviewed for their design basis function during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events. These insights will provide the IDP a means to consider potential impacts of seismic events in the categorization process.

If the MNS seismic hazard changes from medium risk (i.e., Tier 2) at some future time, prior NRC approval, under 10 CFR 50.90, will be requested if the MNS feedback process determines that a process different from the proposed alternative seismic approach is warranted for seismic risk consideration in categorization under 10 CFR 50.69. After receiving NRC approval, Duke Energy will follow its categorization review and adjustment process to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e) and the EPRI 3002017583 SSC categorization criteria for the updated Tier. This includes use of the Duke Energy corrective action process.

If the seismic hazard is reduced such that it meets the criteria for Tier 1 in EPRI 3002017583, Duke Energy will implement the following process.

a. For previously completed system categorizations, Duke Energy may review the categorization results to determine if use of the criteria in EPRI 3002017583 Section 2.2, "Tier 1 - Low Seismic Hazard / High Seismic Margin Sites" would lead to categorization changes. If changes are warranted, they will be implemented through the Duke Energy design control and corrective action programs and NEI 00-04, Section 12.

b. Seismic considerations for subsequent system categorization activities will be performed in accordance with the guidance in EPRI 3002017583 Section 2.2, "Tier 1 - Low Seismic Hazard / High Seismic Margin Sites."

U.S. Nuclear Regulatory Commission Page 19 RA-18-0090 If the seismic hazard increases to the degree that a SPRA becomes necessary to demonstrate adequate seismic safety, Duke Energy will implement the following process following completion of the SPRA, including adequate closure of Peer Review Findings and Observations.

a. For previously completed system categorizations, Duke Energy will review the categorization results using the SPRA insights as prescribed in NEI 00-04 Section 5.3, Seismic Assessment and Section 5.6, "Integral Assessment. If changes are warranted, they will be implemented through the Duke Energy design control and corrective action programs and NEI 00-04 Section 12.

b. Seismic considerations for subsequent system categorization activities will follow the guidance in NEI 00-04, as recommended in EPRI 3002017583 Section 2.4, "Tier 3 - High Seismic Hazard / Low Seismic Margin Sites".

Historical Seismic References for MNS The MNS GMRS and SSE curves from the seismic hazard and screening response are shown in Figure 1, as replicated from the seismic hazard and screening report (Reference 35). The NRC's Staff assessment of the MNS seismic hazard and screening response is documented in Reference 37.

Section 1.1.3 of Reference 17 cites various post-Fukushima seismic reviews performed for the U.S. fleet of nuclear power plants. For MNS, the specific seismic reviews prepared by the licensee and the NRC's staff assessments are provided here. These licensee documents were submitted under oath and affirmation to the NRC.

1. Near-Term Task Force (NTTF) Recommendation 2.1 Seismic Hazard Screening (References 35 and 37).

2. NTTF Recommendation 2.1 Spent Fuel Pool assessment (References 44 and 45).

3. NTTF Recommendation 2.3 Seismic Walkdowns (References 46, 47, and 48).

4. NTTF Recommendation 4.2 Seismic Mitigation Strategy Assessment (S-MSA)

(References 49 and 50).

The following additional post-Fukushima seismic reviews were performed for MNS:

5. NTTF Recommendation 2.1 Expedited Seismic Evaluation Process (ESEP)

(References 41, 42, 43 and 51).

6. NTTF Recommendation 2.1 Seismic High Frequency Evaluation (References 38 and 39).

Technical Information Incorporated by Reference Duke Energy will follow the same alternative seismic approach in the 10 CFR 50.69 categorization process for MNS as the approach that was approved by the NRC staff for LaSalle County Station (Reference 52) with two exceptions:

The MNS LAR cites EPRI Report 3002017583 as applicable to the submittal. The citation for EPRI Report 3002017583 is ADAMS Accession No. ML21082A170. Additionally, the

U.S. Nuclear Regulatory Commission Page 20 RA-18-0090 discussion above cites mark-ups to EPRI Report 3002012988 that were submitted with LaSalle 10 CFR 50.69 LAR RAI responses dated October 16, 2020 (Reference 18) and January 22, 2021 (Reference 19).

The MNS LAR also incorporates the following additional LaSalle 10 CFR 50.69 LAR RAI response (Reference 36) that does not include any mark-ups to EPRI Report 3002017583 but addresses process issues associated with the proposed alternative seismic approach, and specifically excludes LaSalle RAI APLC 50.69-RAI No. 12 that addresses a non-seismic topic (external events).

Summary Based on the above, the Summary from Section 2.3.3 of Reference 17 applies to MNS; namely, MNS is a Tier 2 plant for which there may be a limited number of unique seismic insights, most likely attributed to the possibility of seismically correlated failures, appropriate for consideration in determining HSS SSCs. References 18, 19, and 36 (excludes RAI APLC 50.69-RAI No. 12 that addresses a non-seismic topic (external events)) are incorporated into this LAR as they provide additional supporting bases for Tier 2 plants.

The special sensitivity study recommended using common cause failures, similar to the approach taken in a FPIE PRA, can identify the appropriate seismic insights to be considered with the other categorization insights by the IDP for the final HSS determinations. Use of the EPRI approach outlined in Reference 17 to assess seismic hazard risk for 10 CFR 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).

3.2.4 Other External Hazards All external hazards were screened for applicability to MNS, except seismic, per a plant-specific evaluation in accordance with Generic Letter (GL) 88-20 (Reference 6) and updated to use the criteria in ASME PRA Standard RA-Sa-2009. Attachment 4 provides a summary of the other external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards.

SSCs credited for screening of external hazards 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. Local Intense Precipitation (LIP), dispositioned in Attachment 4, credits the SSCs noted in the table below.

Table 3-2: McGuire Flood Protection Barriers for the LIP Event Unit Elevation Door Number Description 1 760 1EXADR1000C Unit 1 SFP to AB 760' Elevation 1 760 1EXADR1000D Unit 1 SFP to AB 760' Elevation 1 760 0XCDDR1004A AB First Aid Room Outside Door 1 760 0XCDDF1011A Shredder/Compactor Area to Waste Shipping

U.S. Nuclear Regulatory Commission Page 21 RA-18-0090 Unit Elevation Door Number Description 1 760 0XCDDRRA10 Shredder/Compactor Area to Waste Shipping Rollup Door 2 760 0XCDDR1024A AB Hot Machine Shop 2 760 0XCDDR1026D AB 760 Corridor to Hot Tool Room Area 2 760 2EXADF1026C Unit 2 SFP to AB 760' Elevation N/A 784 0XCDDR1255 AB 784' Corridor North End N/A 784 0XCDDF0415A AB 784' Corridor to Central Stairwell 3.2.5 Low Power and Shutdown Consistent with NEI 00-04, the MNS categorization process will use the shutdown safety management plan described in NUMARC 91-06 (Reference 3) 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 NUMARC 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 NEI 00-04 will be considered preliminary HSS.

3.2.6 PRA Maintenance and Updates The Duke Energy risk management process ensures that the applicable PRA models used in this application continues to reflect the as-built and as-operated plant for each of the MNS 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 two refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

In addition, Duke Energy 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,

U.S. Nuclear Regulatory Commission Page 22 RA-18-0090 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 discussed in Section 3.3 of this enclosure.

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.

In the overall risk sensitivity studies, Duke Energy will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference 4. Consistent with the NEI 00-04 guidance, Duke Energy 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 8) and Section 3.1.1 of EPRI TR-1016737 (Reference 9). 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.

Each PRA element notebook was 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 10 CFR 50.69 application in accordance with RG 1.200 Revision 2.

Key MNS PRA model specific assumptions and sources of uncertainty for this application were identified and dispositioned in Attachment 6. The conclusion of this review is that no additional sensitivity analyses are required to address MNS 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 have been assessed against RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 7) consistent with NRC RIS 2007-06.

U.S. Nuclear Regulatory Commission Page 23 RA-18-0090 Internal Events The MNS Units 1 and 2 Internal Events PRA model peer review was performed in June 2015 against ASME/ANS PRA Standard RA-Sa-2009 (Reference 10), RG 1.200 Revision 2 (Reference 7), and NEI 05-04 (Reference 53).

Resolved findings were reviewed and closed in February 2016 using the process documented in the draft of Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) published at the time of the review. Subsequently, the finding closure review was reperformed in May 2019 to the approved process documented in Appendix X to NEI 05-04/07-12/12-13 (References 11 and 12). A subsequent finding closure review was conducted in November 2021 where resolved findings were reviewed and closed using the process documented in NEI 17-07 Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard (Reference 54).

The results of these reviews have been documented and are available for NRC audit.

In conclusion, all the finding level F&Os have been closed, and all associated supporting requirements (SRs) are now judged to be met at Capability Category II or higher. There are no PRA upgrades that have not been peer reviewed.

Large Early Release Frequency (LERF)

The MNS Units 1 and 2 LERF PRA model peer review was performed in December 2012 against ASME/ANS PRA Standard RA-Sa-2009 (Reference 10), RG 1.200 Revision 2 (Reference 7), and NEI 05-04 (Reference 53).

Resolved findings were reviewed and closed in November 2018 using the process documented in Appendix X to NEI 05-04/07-12/12-13 (References 11 and 12). A subsequent finding closure review was conducted in June 2022 where resolved findings were reviewed and closed using the process documented in NEI 17-07 (Reference 54).

The results of these reviews have been documented and are available for NRC audit.

In conclusion, all the finding level F&Os have been closed, and all associated SRs are now judged to be met at Capability Category II or higher. There are no PRA upgrades that have not been peer reviewed.

Internal Flooding The MNS Units 1 and 2 Internal Flooding PRA model peer review was performed in September 2011 against ASME/ANS PRA Standard RA-Sa-2009 (Reference 10), RG 1.200 Revision 2 (Reference 7), and NEI 05-04 (Reference 53).

A finding closure review was conducted on the Internal Flooding PRA model in November 2018 where resolved findings were reviewed and closed using the process documented in Appendix X to NEI 05-04/07-12/12-13 (References 11 and 12). A subsequent finding closure review was conducted in June 2022 where resolved findings were reviewed and closed using the process documented in NEI 17-07 (Reference 54).

The results of these reviews have been documented and are available for NRC audit.

U.S. Nuclear Regulatory Commission Page 24 RA-18-0090 In conclusion, all the finding level F&Os have been closed, and all associated SRs are now judged to be met at Capability Category II or higher. There are no PRA upgrades that have not been peer reviewed.

Fire The MNS Units 1 and 2 Fire PRA model peer review was performed in 2010 against ASME/ANS PRA Standard RA-Sa-2009 (Reference 10), RG 1.200 Revision 2 (Reference 7),

and NEI 07-12 (Reference 55). In 2019, 2020, 2021, and 2022 Focused Scope Peer Reviews (FSPRs) were performed to address various PRA upgrades where all of the FSPRs were performed against ASME/ANS PRA Standard RA-Sa-2009 (Reference 10), RG 1.200 Revision 2 (Reference 7), and NEI 07-12 (Reference 56) or NEI 17-07 (Reference 54), as applicable at the time of the review.

Finding closure reviews were conducted on the Fire PRA model in January 2019 and December 2020 where resolved findings were reviewed and closed using the process documented in Appendix X to NEI 05-04/07-12/12-13 (References 11 and 12). Subsequent closure reviews were conducted on the Fire PRA model in November 2021 and September 2022 where resolved findings were reviewed and closed using the process documented in NEI 17-07 (Reference 54). In some instances, a FSPR and F&O closure review were conducted in parallel, however, in each instance, the scope for the F&O closure did not include any findings generated from the parallel FSPR. The following list provides a summary of the scope for each review and the detailed reports for each review are available for NRC audit:

x January 2019 FSPR: Assesses a model upgrade to use Human Reliability Analysis (HRA) Calculator Software for Fire PRA Human Failure Event (HFE) Analysis x January 2019 F&O Closure: This closure review was performed in parallel to the January 2019 FSPR and does not assess findings generated in the parallel FSPR x December 2020 FSPR: Assesses a model upgrade to use the Obstructed Radiation Method x December 2020 F&O Closure: Assesses two findings generated in the January 2019 FSPR on HRA x November 2021 FSPR: Assesses two newly applicable SRs and one SR previously assessed at Capability Category I with no open finding x November 2021 F&O Closure: Assesses select findings from 2020 FSPR x August 2022 FSPR: Assesses a newly applicable SR and a model upgrade to use a quantitative analysis for fire impacts to structural steel x September 2022 F&O Closure: Assesses closure of the finding from the August 2022 FSPR The results of these reviews have been documented and are available for NRC audit.

U.S. Nuclear Regulatory Commission Page 25 RA-18-0090 In conclusion, all the finding level F&Os have been closed, and all associated SRs are now judged to be met at Capability Category II or higher. There are no PRA upgrades that have not been peer reviewed. reflects zero open finding level F&Os for MNS PRA models. The attachments identified above demonstrate that the PRA is of sufficient quality and level of detail to support the categorization process 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 MNS 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04. The overall risk evaluation process described in the NEI 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 core damage frequency (CDF) and 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.

The performance monitoring process is described in Duke Energys 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. Station personnel from Engineering, Operations, Risk Management, Regulatory Affairs, and others have responsibilities for preparing and conducting various performance monitoring tasks that feed into this process. The intent of the performance monitoring reviews is to discover trends in component reliability, to help catch and reverse negative performance trends and take corrective action if necessary.

To more specifically address the feedback and adjustment (i.e., performance monitoring) process as it pertains to the proposed MNS Tier 2 approach discussed in Section 3.2.3, implementation of the Duke Energy 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).

Duke Energy 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

U.S. Nuclear Regulatory Commission Page 26 RA-18-0090 categorization process will be identified and addressed through the problem identification and corrective action program, including seismic-related issues.

The Duke Energy 10 CFR 50.69 program requires that SCDs cannot be approved by the IDP until the panels 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.

Scheduled periodic reviews are completed at least once every two refueling cycles in accordance with Duke Energy procedures and 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:

x A review of plant modifications since the last review that could impact the SSC categorization x A review of plant specific operating experience that could impact the SSC categorization, x A review of the impact of the updated risk information on the categorization process results x A review of the importance measures used for screening in the categorization process.

x An update of the risk sensitivity study performed for the categorization x Input from Regulatory Affairs and Operations regarding changes that may affect the bases for the categorization results.

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.

x 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."

x 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.

x Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, April 2015.

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

U.S. Nuclear Regulatory Commission Page 27 RA-18-0090 The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 No Significant Hazards Consideration Determination Analysis Duke Energy Carolinas, LLC (Duke Energy) proposes to modify the licensing basis for McGuire Nuclear Station, Units 1 and 2, 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 significance resulting in improved plant safety.

Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment 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 Structures, Systems and Components (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

U.S. Nuclear Regulatory Commission Page 28 RA-18-0090 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 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, Duke Energy 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 Conclusions 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.

U.S. Nuclear Regulatory Commission Page 29 RA-18-0090

6. REFERENCES

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

2. 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.

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

4. NRC letter to Southern Nuclear Operating Company, Inc., Vogtle Electric Generating Plant Units 1 and 2 - Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC Nos. ME9472 and ME94473), dated December 17, 2014 (ADAMS Accession No. ML14237A034)

5. NRC letter 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), dated April 22, 2009 (ADAMS Accession No. ML090930246).

6. Generic Letter 88-20, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4, USNRC, June 1991.

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

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

9. EPRI TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, December 2008

10. ASME/ANS RA-Sa-2009, Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, dated February 2009

11. NEI Letter to NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os), February 21, 2017 (ADAMS Accession No. ML17086A431).

12. NRC Letter to Mr. Greg Krueger (NEI), U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), May 3, 2017 (ADAMS Accession No. ML17079A427).

13. NRC Letter to Mr. Oliver Martinez, U.S. Nuclear Regulatory Commission (NRC)

Comments on Addenda to a Current ANS: ASME RA-SB - 20XX, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment For Nuclear Power Plant Applications, dated July 6, 2011 (ADAMS Accession No. ML111720076).

14. NRC Letter to Duke Energy, McGuire Nuclear Station, Units 1 AND 2, Issuance of Amendments Regarding Revision of the Technical Specifications to Relocate Specific Surveillance Frequencies to a Licensee-Controlled Program Using a Risk-Informed Justification (TSTF-425), dated March 29, 2011 (ADAMS Accession No. ML110680357).

U.S. Nuclear Regulatory Commission Page 30 RA-18-0090

15. Letter from NRC to Duke Energy, McGuire Nuclear Station, Units 1 and 2 - Issuance of Amendments Regarding National Fire Protection Association Standard (NFPA) 805, dated December 6, 2016 (ADAMS Accession No. ML16077A135).

16. Letter from NRC to Duke Energy, Review of McGuire Nuclear Station, Units 1 and 2 -

Individual Plant Examination of External Events Submittal, dated February 16, 1999 (ADAMS Accession No. ML20203J043).

17. Electric Power Research Institute (EPRI) 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, February 2020 (ADAMS Accession No. ML21082A170).

18. Exelon Generation Company, LLC Letter to NRC, Response to Request for Additional Information Regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors," dated October 16, 2020 (ADAMS Accession No. ML20290A791).

19. Exelon Generation Company, LLC Letter to NRC, "Response to Request for Additional Information Regarding the License Amendment Request to Adopt 10 CFR 50.69," dated January 22, 2021 (ADAMS Accession No. ML21022A130).

20. NRC letter to all Power Reactor Licensees, "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 (ADAMS Accession No. ML12053A340)

21. Exelon Generation Company, LLC Letter to NRC, Seismic Probabilistic Risk Assessment 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," dated August 28, 2018 (ADAMS Accession No. ML18240A065).

22. NRC Letter to Exelon Generation Company, LLC, 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 (ADAMS Accession No. ML19053A469)

23. NRC Letter to Exelon Generation Company, LLC, 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 (ADAMS Accession No. ML19248C756)

24. Southern Nuclear Operating Company, Inc. Letter to NRC, Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process," June 22, 2017 (ADAMS Accession No. ML17173A875)

25. NRC Letter to Southern Nuclear Operating Company, Inc., 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," August 10, 2018 (ADAMS Accession No. ML18180A062)

U.S. Nuclear Regulatory Commission Page 31 RA-18-0090

26. Tennessee Valley Authority Letter to NRC, Seismic Probabilistic Risk Assessment for Watts Bar Nuclear Plant, Units 1 and 2, Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," June 30, 2017 (ADAMS Accession No. ML17181A485)

27. Tennessee Valley Authority Letter to NRC, Tennessee Valley Authority (TVA) - Watts Bar Seismic Probabilistic Risk Assessment Supplemental Information, April 10, 2018 (ADAMS Accession No. ML18100A966)

28. NRC Letter 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 (ADAMS Accession No. ML18115A138)

29. NRC Letter to Tennessee 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 (ADAMS Accession No. ML20076A194)

30. EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin", Revision 1, August 1991

31. Exelon Generation Company, LLC, letter to NRC, "Supplemental Information to Support Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power plants," June 6, 2018 (ADAMS Accession No. ML18157A260)

32. Southern Nuclear Operating Company, Inc. letter to NRC, "Vogtle Electric Generating Plant, Units 1 & 2, License Amendment Request to Incorporate Seismic Probabilistic Risk Assessment into 10 CFR 50.69 Categorization Process, Response to Request for Additional Information (RAIs 4-11)," February 21, 2018 (ADAMS Accession No. ML18052B342)

33. Tennessee Valley Authority Letter to 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," November 29, 2018 (ADAMS Accession No. ML18334A363)

34. MCC-1535.00-00-0252 Revision 1, MNS 50.69 and TSTF-505 LAR Support Calculation

35. Duke Energy Letter to NRC, "Seismic Hazard and Screening Report (CEUS Sites),

Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations Section 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,"

(ADAMS Accession No. ML14098A421), dated March 20, 2014.

36. Exelon Generation Company, LLC Letter to NRC, "Response to Request for Additional Information Regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power plants," dated October 1, 2020 (ADAMS Accession No. ML20275A292).

U.S. Nuclear Regulatory Commission Page 32 RA-18-0090

37. NRC Letter to Duke Energy Carolinas, LLC, Catawba Nuclear Station, Units 1 and 2, and McGuire Nuclear Station, Units 1 and 2, Screening and Prioritization Results Regarding Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-ichi Accident," (ADAMS Accession No. ML16344A313), dated December 22, 2016.

38. Duke Energy Letter to NRC, High Frequency Supplement to Seismic Hazard Screening Report, Response 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, (ADAMS Accession No. ML17230A088), dated August 10, 2017.

39. NRC Letter to Duke Energy Carolinas, LLC, McGuire Nuclear Station, Units 1 and 2 - Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 (ADAMS Accession No. ML17320A770), dated November 20, 2017.

40. Duke Energy Letter to NRC, "Supplemental Information Regarding Reevaluated Seismic Hazard Screening and Prioritization Results - Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," (ADAMS Accession No. ML16295A342), dated October 20, 2016.

41. Duke Energy Letter to NRC, Expedited Seismic Evaluation Process (ESEP) Report (CEUS Sites), Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, (ADAMS Accession No. ML15005A085), dated December 17, 2014.

42. Duke Energy Letter to NRC, Expedited Seismic Evaluation Process (ESEP) Report (CEUS Sites), Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, (ADAMS Accession No. ML15313A153), dated October 8, 2015.

43. Duke Energy Letter to NRC, Expedited Seismic Evaluation Process (ESEP) Closeout, Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, ADAMS Accession No. ML16041A173), dated February 4, 2016.

44. Duke Energy Letter to NRC, Spent Fuel Pool Evaluation Supplemental 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, (ADAMS Accession No. ML16236A074), dated August 18, 2016.

45. NRC Letter to Duke Energy Carolinas, LLC, McGuire Nuclear Station, Units 1 and 2 - Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1, (ADAMS Accession No. ML16237A354), dated August 31, 2016.

46. Duke Energy Letter to NRC, Seismic Walkdown Information Requested by NRC Letter, 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; dated March 12, 2012, (ADAMS Accession No. ML13003A339), dated November 26, 2012.

U.S. Nuclear Regulatory Commission Page 33 RA-18-0090

47. Duke Energy Letter to NRC, Response to Request for Additional Information Regarding the Seismic Hazard Walkdowns Associated With Near-Term Task Force Recommendation 2.3, Seismic Walkdowns, (ADAMS Accession No. ML13338A171), dated November 26, 2013.

48. NRC Letter to Duke Energy Carolinas, LLC, McGuire Nuclear Station, Unit 1- 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, (ADAMS Accession No. ML14114A305), dated May 8, 2014.

49. Duke Energy Letter to NRC, McGuire Nuclear Station (MNS) Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Appendix H, Revision 2, H.4.4 Path 4: GMRS < 2xSSE, (ADAMS Accession No. ML17233A167), dated August 10, 2017.

50. NRC Letter to Duke Energy Carolinas, LLC, McGuire Nuclear Station, Units 1 and 2 - Staff Review of Mitigating Strategies Assessment Report of the Impact of the Re-Evaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter, (ADAMS Accession No. ML17349A991), dated December 21, 2017.

51. NRC Letter to Duke Energy Carolinas, LLC, McGuire Nuclear Station, Units 1 and 2 - Staff Review of Interim Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1, (ADAMS Accession No. ML16072A038),

dated March 17, 2016.

52. NRC letter to Exelon Generation Company, LLC, Lasalle County Station, Unit Nos. 1 And 2 - Issuance of Amendment Nos. 249 And 235 Related to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, dated May 27, 2021 (ADAMS Accession No. ML21082A422).

53. NEI 05-04 Revision 3, Process for Performing Internal Events Peer Reviews Using the ASME/ANS PRA Standard, November 2009.

54. NEI 17-07 Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, August 2019.

55. NEI 07-12 Revision 0, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines," Nuclear Energy Institute, November 2008.

56. NEI 07-12 Revision 1, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines," Nuclear Energy Institute, June 2010.

57. Electric Power Research Institute (EPRI) 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, July 2018.

58. NRC Letter 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. (ADAMS Accession No. ML19330D909)

59. Exelon Generation Company, LLC Letter to NRC, Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-505, Revision 2, and 10 CFR 50.69, November 24, 2020. (ADAMS Accession No. ML20329A433).

U.S. Nuclear Regulatory Commission RA-18-0090 ATTACHMENT 1 LIST OF CATEGORIZATION PREREQUISITES

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U.S. Nuclear Regulatory Commission Page 1 RA-18-0090 Duke Energy 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.

x Integrated Decision-Making Panel (IDP) member qualification requirements.

x 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 of the enclosure for this license amendment request). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting, an LSS function are categorized as preliminary LSS.

x 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.

x 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.

x Review by the IDP. The categorization results are presented to the IDP for review and approval. The IDP reviews the categorization results and makes the final determination on the safety significance of system functions and components.

x 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.

x Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized.

x Documentation requirements per Section 3.1.1 of the enclosure.

U.S. Nuclear Regulatory Commission RA-18-0090 ATTACHMENT 2 DESCRIPTION OF PRA MODELS USED IN CATEGORIZATION

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U.S. Nuclear Regulatory Commission Page 1 RA-18-0090 Unit 1 CDF & LERF Baseline CDF Baseline LERF Source Contribution Source Contribution Internal Events Internal Events 3.14E-06 3.96E-07 PRA PRA Internal Internal 4.86E-06 5.89E-07 Flooding PRA Flooding PRA Fire PRA 3.37E-05 Fire PRA 5.12E-06 Total CDF 4.17E-05 Total LERF 6.11E-06 Unit 2 CDF & LERF Baseline CDF Baseline LERF Source Contribution Source Contribution Internal Events Internal Events 3.16E-06 4.20E-07 PRA PRA Internal Internal 6.38E-06 6.17E-07 Flooding PRA Flooding PRA Fire PRA 4.06E-05 Fire PRA 5.01E-06 Total CDF 5.01E-05 Total LERF 6.05E-06

U.S. Nuclear Regulatory Commission RA-18-0090 ATTACHMENT 3 DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS

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U.S. Nuclear Regulatory Commission Page 1 RA-18-0090 The McGuire Nuclear Station (MNS), Units 1 and 2 Internal Events, Internal Flood, and Fire PRA models have zero open finding level Facts and Observations (F&Os). As such, is not applicable to MNS.

U.S. Nuclear Regulatory Commission RA-18-0090 ATTACHMENT 4 EXTERNAL HAZARDS SCREENING

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U.S. Nuclear Regulatory Commission Page 1 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 Per the IPEEE (Reference 1), an assessment of aircraft impact risk was performed with the total crash probability listed as 1.3E-08/yr.

The assessment was updated using recent air traffic data using the guidance provided in NUREG 0800 Section 3.5.1.6 (Reference 2). The primary change that has occurred since the conduct of the IPEEE analyses is the increase in air traffic, particularly commercial air traffic that uses the Charlotte-Douglas International Airport Aircraft Impact Y PS4 (CLT).

The analysis resulted in an updated value of the annual probability of an aircraft crash onto the MNS site of 3.2E-8, which is a factor of 3 below the risk criteria specified in the SRP for probability of aircraft accidents that could result in releases that exceed 10 CFR 100 limits of less than 1E-7 per year. Additionally, the crash frequency is much less than the CDF screening criteria of 1E-6 per year (PS4).

Based on this review, the Aircraft Impact hazard is considered to be negligible.

Per the IPEEE (Reference 1), there are no mountains in the vicinity of McGuire from which a significant avalanche could Avalanche Y C3 be generated.

Based on this review, the Avalanche hazard is considered to be negligible.

This hazard is slow to develop and can Biological Events Y C5 be identified via monitoring and

U.S. Nuclear Regulatory Commission Page 2 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 managed via standard maintenance process. Actions committed to and completed by MNS in response to Generic Letter 89-13 provide on-going control of biological hazards. These include performance of periodic maintenance work orders to inspect the intake structures, perform flow balance/testing, periodic flushing, and heat exchanger cleaning.

Based on this review, the Biological Events hazard is considered to be negligible.

Per the IPEEE (Reference 1), McGuire is located more than 150 miles from the nearest coastal area. However, to protect the lake edge from erosion, the yard areas subjected to waves are protected by riprap underlain by a thick Coastal Erosion Y C1 subgrade of filter material. Therefore, lake edge erosion will not be a significant problem.

Based on this review, the Coastal Erosion hazard is considered to be negligible.

Drought is a slowly developing hazard allowing time for orderly plant Drought Y C1 reductions, including shutdowns.

Based on this review, the Drought hazard is considered to be negligible.

Per the Flood Hazard Reevaluation I

External Flood Y C1 Report (FHRR - Reference 4), external I I

U.S. Nuclear Regulatory Commission Page 3 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 flooding mechanisms exceeding the Current Design Basis (CDB) that could pose a potential challenge to MNS key safety functions were:

Local Intense Precipitation (LIP)

Flooding in Streams and Rivers (referred to as Flooding in Reservoirs in MNS FHRR)

Failure of Dams

Probable Maximum Storm Surge and Seiche/Wind Wave Runup To mitigate worst case flood scenarios for mechanisms other than LIP (henceforth referred to as Combined Effects (CE) flooding), permanent concrete protective barriers on the north embankment were installed to raise the flood protection levels at the site to 779 ft. This prevents water from the CE flood encroaching on the Auxiliary Building (AB) and site grade. This modification is permanent, passive protection that does not require any human actions to keep the site dry.

There is 0.46 ft of available physical margin (APM) present with the barriers installed.

The LIP flood causing mechanism produces a maximum water surface elevation of 761.1 ft (1.1 ft of standing water for 2.5 hrs) around the AB which houses all SSCs related to maintaining key safety functions. For this LIP flood

U.S. Nuclear Regulatory Commission Page 4 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 mechanism, site procedures require installation of temporary, engineered flood barriers at several locations around the AB.

The site begins preparations to install the flood barriers approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before the arrival of the storm.

The barriers are engineered for rapid deployment and ample recovery time is available should troubleshooting or reinstallation be required. The barriers are rated to protect AB doors from 30 (2.5 ft) of water and only 0.6 ft of water has been postulated due to a LIP event.

With the barriers in place, all SSCs related to KSFs are maintained free from flood waters throughout the event with adequate available physical margin.

The flood protection door barriers will be categorized as high safety significant (HSS) should categorization of the system be completed.

Based on this review, the External Flood hazard is considered to be negligible.

The total CDF for High Wind (HW) hazards is 3.0E-6/yr and 3.1E-6/yr for Units 1 and 2, respectively. LERF is C4 approximately 1.1E-7/yr for both units.

Extreme Winds While these values are above the Y numerical screening criteria of 1E-6/yr and Tornadoes PS4 and 1E-7/yr for CDF and LERF, a large percentage of CDF and LERF (approximately 75% and 65%,

respectively) include no wind induced failures except for offsite power; they

U.S. Nuclear Regulatory Commission Page 5 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 are wind-induced LOOP events with random equipment and operator failures and failure to recover offsite power.

Additionally, nearly all of these scenarios are associated with wind speeds in the F1 - F2 range (less than or equal to 157 mph).

The CDF and LERF due to scenarios/sequences involving wind-induced failures, either wind pressure or missiles, are estimated to be approximately 8E-7/yr and 4E-8/yr, respectively. Thus, the average maintenance CDF and LERF for these HW Failure Only scenarios are less than 1E-6/yr and 1E-7/yr.

Note that NEI 00-04 Figure 5-6 requires that, as part of system categorization, an evaluation be conducted to determine if there are components that participate in screened scenarios and whose failure would result in an unscreened scenario. Such SSCs are required to be high safety significant (HSS) in the categorization process.

These HSS SSCs for the MNS High Winds hazard include Category I Structures which are already considered HSS. During categorization of systems, MNS will use NEI 00-04 Figure 5-6 as appropriate to determine the safety significance of components in systems being categorized.

Therefore, all HW hazards screened via the internal events weather-related LOOP sequences (C4 and PS4).

U.S. Nuclear Regulatory Commission Page 6 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 Based on this review, the Extreme Winds and Tornadoes hazard is considered to be negligible.

Per the IPEEE (Reference 1), accident data involving surface vehicles or aircraft would include the effects of fog.

Per the UFSAR Section 2.3.2.3 (Reference 5), consideration has been given to possible environmental effects associated with heat dissipation from Fog Y C1 the cooling pond (Lake Norman, vicinity of McGuire Nuclear Station). A review of the literature and operating experience to date would suggest that effects of fogging and icing are minimal for the properly designed cooling pond.

Based on this review, the Fog hazard is considered to be negligible.

Per the IPEEE (Reference 1), bush and local forest fires are handled by the local fire department. Such fires are not considered to have any impact on the Forest Fire Y C1 station because the site is cleared and the fire cannot propagate to station buildings or equipment.

U.S. Nuclear Regulatory Commission Page 7 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 Per the UFSAR Section 2.2.3 (Reference 5), the only potential fire hazard in the plant vicinity is a brush fire. The plant fire protection system is adequate to prevent any possible damage from a fire due to this origin.

Based on this review, the Forest Fire hazard is considered to be negligible.

Per the IPEEE (Reference 1), both the Reactor Building and the Auxiliary Building are designed for a combination Frost Y C1 of snow, ice, and rain.

Based on this review, the Frost hazard is considered to be negligible.

Per the IPEEE (Reference 1), both the Reactor Building and the Auxiliary Building are designed for a combination of snow, ice, and rain.

C1 In addition, the principal effects of such Hail Y events would be to cause a loss of off-C4 site power, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for McGuire.

Based on this review, the Hail hazard is considered to be negligible.

Per the IPEEE (Reference 1), the effect of high summer temperatures at C1 High Summer McGuire is insignificant because there Y

Temperature are upstream dams that provide water C4 level control on Lake Norman.

U.S. Nuclear Regulatory Commission Page 8 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 Technical Specification Surveillance Requirement (SR) 3.7.8.2 verifies that the Nuclear Service Water System is available to cool the CCW System to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident. The SR verifies that the average water temperature of the Standby Nuclear Service Water Pond is 82°F at an elevation of 722 ft (level must be 739.5 ft per SR 3.7.8.1) else the plant be shut down.

The SR is modified by a Note that states the Surveillance is only required to be performed during the months of July, August, and September. During other months, the ambient temperature is below the surveillance limit.

In addition, the principal effects of such events would be to cause a loss of off-site power, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for McGuire.

Based on this review, the High Summer Temperature hazard is considered to be negligible.

Per the IPEEE (Reference 1), McGuire is located more than 150 miles from the High Tide Y C4 nearest coastal area.

U.S. Nuclear Regulatory Commission Page 9 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 I

Based on this review, the High Tide hazard is considered to be negligible.

I I See Extreme Winds or Tornados and External Flood / Intense Precipitation.

External Flood effects from hurricanes Hurricane are accounted for in the Combined Y C4 (Tropical Cyclone) Effects (CE) hazard analysis.

Based on this review, the Hurricane (Tropical Cyclone) hazard is considered to be negligible.

Per the IPEEE (Reference 1), Both the Reactor Building and the Auxiliary Building are designed for ice.

Per UFSAR (Reference 5) Section 2.4.7, ice flooding is not applicable to C1 the site area.

Ice Cover Y C4 In addition, the principal effects of such events would be to cause a loss of off-site power, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for McGuire.

Based on this review, the Ice Cover hazard is considered to be negligible.

Per the IPEEE (Reference 1), there are no military or industrial facilities within a Industrial or C1 5-mile radius of the plant.

Military Facility Y Accident C3 Per UFSAR Section 2.2 (Reference 5),

military and transportation facilities are nearly non-existent and only a few

U.S. Nuclear Regulatory Commission Page 10 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 industrial facilities are located in the vicinity of McGuire. The few facilities that do exist have no effect on the McGuire Nuclear Station nor will McGuire Nuclear Station have any effect on the existing facilities.

Based on this review, the Industrial or Military Facility Accident hazard is considered to be negligible.

The McGuire Internal Events and Internal Flood N/A N/A Internal Flood PRA model addresses risk from internal flood events.

Internal Fire N/A N/A The McGuire Internal Fire PRA model addresses risk from internal fires.

Per the IPEEE (Reference 1), landslides are considered an insignificant hazard at McGuire. The Standby Nuclear Service Water Pond (SNSWP) dam is the only natural or man made slope which, upon Landslide Y C1 failure, would prevent safe shutdown of the plant. Therefore, the SNSWP was statically designed for stability under all loading conditions Based on this review, the Landslide hazard is considered to be negligible.

Per the IPEEE (Reference 1), the most probable effect of lightning is the loss of off-site power due to a strike in the Lightning Y C4 switchyard. These occurrences are accounted for in the loss of off-site power initiating event frequency.

U.S. Nuclear Regulatory Commission Page 11 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 I

Based on this review, the Lightning hazard is considered to be negligible.

I I Per the IPEEE (Reference 1), the effect of low lake level, or low river water level at McGuire is insignificant because there are upstream dams that provide water level control on Lake Norman.

Due to normal regulation of lake level Low Lake or River and the extended time available before Y C1 Water Level minimum water level, sufficient time would be available to respond to this hazard.

Based on this review, the Low Lake or River Water Level hazard is considered to be negligible.

Per the IPEEE (Reference 1), the Reactor Building and the Auxiliary Building are designed for a combination of snow and ice. These hazards are commensurate with low winter C1 temperatures.

Low Winter Y

Temperature In addition, low winter temperatures C4 causing failure of instruments are included in the plant trip frequency data.

Based on this review, the Low Winter Temperature hazard is considered to be negligible.

U.S. Nuclear Regulatory Commission Page 12 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 Per the IPEEE (Reference 1), this event has significantly lower frequency than other events with similar uncertainties.

The frequency of a meteor or satellite strike is judged to be so low as to make Meteorite/Satellite Y PS4 the risk impact from such events Strikes insignificant.

Based on this review, the Meteorite/Satellite Strikes hazard is considered to be negligible.

Per the IPEEE (Reference 1), gas pipeline maps of the area around the McGuire plant site were reviewed and indicated that there were no changes to the original PRA screening information as contained in the FSAR.

Per the FSAR Section 2.2.3 (Reference 5), there are two gas pipelines: one 36-inch diameter and one 42-inch diameter located one mile south of the plant. The Pipeline Accident Y C3 consequences a rupture of the 42-inch gas pipeline rupture was evaluated.

The evaluation included the potential effects of the gas at the plant, an unconfined in-air explosion, and surface blast at the point of rupture.

The evaluation found the effects of gas at the plant were well below the flammability threshold. The unconfined in-air explosion and surface blast effects only resulted in a worst-case

U.S. Nuclear Regulatory Commission Page 13 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 overpressure of 1.3 to 1.8 psi at the plant, which is considered minor.

Based on this review, the Pipeline Accident hazard is considered to be negligible.

This hazard is discussed under External Flooding.

Precipitation, See External Flood.

Y C1 Intense Based on this review, the Intense Precipitation hazard is considered to be negligible.

Per the IPEEE (Reference 1), potential hazards from the storage of toxic material on-site is minimal.

The FSAR Section 2.1.4 (Reference 5) states that no large quantities of caustic or flammable material will be stored on site.

Release of MNS updated its Toxic Gas evaluation Chemicals from Y C1 in July 2022 (Reference 6) to evaluate Onsite Storage onsite and offsite chemical hazards in accordance with Regulatory Guide 1.78, Rev. 1 (Reference 7).

The evaluation considered potential onsite and offsite stationary and mobile hazardous chemical sources that could pose a threat to control room habitability upon release within 5 miles of MNS.

U.S. Nuclear Regulatory Commission Page 14 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 The evaluation concluded that there are no toxic gas hazardous chemical threats to control room habitability.

Based on this review, the Release of Chemicals from Onsite Storage hazard is considered to be negligible.

Per the IPEEE (Reference 1), no present means exist to divert or reroute the river flow through the dams other than insignificant amounts of water used for municipal supply.

Per UFSAR Section 2.4.9 (Reference 5), there are five reservoirs on the Catawba River upstream of Cowans Ford Dam, all of which have operating hydroelectric power plants located on River Diversion Y C1 them. Since Duke owns and controls the levels of each reservoir above the site of McGuire Nuclear Station, any upstream diversion or rerouting of the source of cooling water is very unlikely to happen. No present means exist to divert or reroute other than minor amounts used for municipal water supply.

Based on this review, the River Diversion hazard is considered to be negligible.

Per the IPEEE (Reference 1), McGuire Sandstorm Y C1 is located more than 150 miles from the nearest area with a large sand deposit.

U.S. Nuclear Regulatory Commission Page 15 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 The likelihood of occurrence is insignificant.

Based on this review, the Sandstorm hazard is considered to be negligible.

Per FSAR Section 2.4.5.2 (Reference 5), Lake Norman, immediately north of the plant, is a relatively new inland lake with no history of surge or seiche Flood.

Seiche Y C1 The Seiche hazard is accounted for in the Combined Effects (CE) hazard analysis for External Flood.

Based on this review, the Seiche hazard is considered to be negligible.

MNS is an EPRI Tier 2 Plant as defined Seismic Activity N/A N/A by Reference 8. See Section 3.2.3 of this LAR.

Per the IPEEE (Reference 1), both the Reactor Building and the Auxiliary Snow Y C1 Building are designed for snow.

Based on this review, the Snow hazard is considered to be negligible.

Per FSAR Section 2.5 (Reference 5),

extensive investigations on soil and rock samples found that subsurface conditions of the site have no adverse Soil Shrink-Swell Y C1 impact on the design, construction, or operation of the station.

Based on this review, the Soil Shrink-Swell hazard is considered to be negligible.

U.S. Nuclear Regulatory Commission Page 16 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 The Storm Surge hazard is accounted for in the Combined Effects (CE) hazard Storm Surge Y C1 analysis for External Flood.

Based on this review, the Storm Surge hazard is considered to be negligible.

Per the IPEEE (Reference 1), leaks from containers of chlorine (used for drinking water purification and sanitary waste treatment) and other potential toxic gas sources were evaluated which found that it is unlikely that leaks from these containers would result in dangerous concentrations in the Control Room.

MNS updated its Toxic Gas evaluation in July 2022 (Reference 6) to evaluate onsite and offsite chemical hazards in accordance with Regulatory Guide 1.78, Toxic Gas Y C1 Rev. 1 (Reference 7).

The evaluation considered potential onsite and offsite stationary and mobile hazardous chemical sources that could pose a threat to control room habitability upon release within 5 miles of MNS.

The evaluation concluded that there are no toxic gas hazardous chemical threats to control room habitability.

See also Release of Chemicals from Onsite Storage.

Based on this review, the Toxic Gas hazard is considered to be negligible.

U.S. Nuclear Regulatory Commission Page 17 RA-18-0090 Screening Result1,3 External Hazard Screened? Screening Comment (Y/N) Criterion2 Per the IPEEE (Reference 1), there are no industries within 5 miles of McGuire which transport or store products harmful to the station.

Per FSAR Section 2.2.2 (Reference 5),

the major transportation corridors within 5 miles of the site N.C. 16, located approximately three miles west of the site and I-77 located approximately five miles east of the site. N.C. 73, located approximately 0.4 miles south of the site, is primarily used by local residents, commuters, and for recreational access to Lake Norman. There are no Transportation manufacturers or suppliers of hazardous Y C3 Accidents materials within 10 miles of the site.

The shipment of hazardous materials is regulated by the U.S. Department of Transportation (USDOT). Based on the USDOT regulations and the proximity of alternate major high-speed highways bypassing the site, the probability of MNS being affected by shipment of hazardous materials is insignificant.

v t e Letter Sequence Request
EPID:L-2023-LLA-0021, TSTF 505 (Open)
Initiation Request, Request Acceptance Administration Questions Answers Results Approval... Other: ML23087A249, ML23094A183, ML24019A139
MONTHYEAR2023-03-27 [Table View]

RA-18-0090, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

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RA-18-0090, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

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