NL-24-0051, License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components

License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components

ML24051A239
Person / Time
Site:	Hatch
Issue date:	02/20/2024
From:	Coleman J Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-24-0051
Download: ML24051A239 (1)

Text

>. Southern Nuclear February 20, 2024 Docket Nos.: 50-321 50-366 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Regulatory Affairs Edwin I. Hatch Nuclear Plant - Units 1 and 2 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5000 NL-24-0051 10 CFR 50.90 10 CFR 50.69 License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Southern Nuclear Operating Company (SNC) is submitting a request for an amendment to renewed facility operating licenses (RFOL) DPR-57 and NPF-5 for Edwin I.

Hatch Nuclear Plant (HNP), Units 1 and 2, respectively.

The proposed amendment would modify the licensing basis to implement a change to the approved voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors." The proposed amendment would incorporate the use of an alternative seismic method, as described in this request, in addition to the peer reviewed, plant-specific HNP seismic probabilistic risk assessment (SPRA) into the previously approved 10 CFR 50.69 categorization process, as allowed by the Nuclear Regulatory Commission endorsed industry guidance.

The Enclosure provides a description and assessment of the proposed change. The Attachment provides the existing RFOL pages marked up to show the proposed change.

SNC requests approval of the proposed license amendment 12 months following acceptance.

The proposed change would be implemented within 90 days of issuance of the amendment.

SNC has concluded that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, "Issuance of amendment."

In accordance with 10 CFR 50.91, SNC is notifying the State of Georgia of this license amendment request by transmitting a copy of this letter, with attachments, to the designated State Official.

U. S. Nuclear Regulatory Commission NL-24-0051 Page 2 This letter contains no NRC commitments. If you have any questions, please contact Ryan Joyce at 205.992.6468.

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

Executed on the 20th day of February 2024.

Respectfully submitted,

~~

Jamie M. Coleman Director, Regulatory Affairs Southern Nuclear Operating Company JMC/RMJ

Enclosure:

Description and Assessment of the Proposed Change

Attachment:

Markups of HNP Units 1 & 2 Renewed Facility Operating Licenses cc:

NRC Regional Administrator, Region II NRR Project Manager - Hatch Senior Resident Inspector - Hatch Director, Environmental Protection Division - State of Georgia RType: CHA02.004

ENCLOSURE Edwin I. Hatch Nuclear Plant - Units 1 and 2 License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components Description and Assessment of the Proposed Change

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change Enclosure Description and Assessment of the Proposed Change TABLE OF CONTENTS

1.

SUMMARY

DESCRIPTION......................................................................................... E-2

2.

DETAILED DESCRIPTION.......................................................................................... E-3 2.1 Current Regulatory Requirements.................................................................... E-3 2.2 Reason for Proposed Change.......................................................................... E-4 2.3 Description of the Proposed Change............................................................... E-4

3.

TECHNICAL EVALUATION......................................................................................... E-5 3.1 Categorization Process Description (10 CFR 50.69(b)(2)(i))............................ E-6 3.1.1 Categorization Process Using Alternative Seismic Method for Tier 2 Sites..................................................................................... E-6 3.2 Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii))............................... E-16 3.3 PRA Review Process Results (10 CFR 50.69(b)(2)(iii))................................. E-17 3.4 Risk Evaluations (10 CFR 50.69(b)(2)(iv))..................................................... E-17 3.5 Feedback and Adjustment Process................................................................ E-17

4.

REGULATORY EVALUATION................................................................................... E-19 4.1 Applicable Regulatory Requirements/Criteria...................................................... E-19 4.2 No Significant Hazards Consideration Analysis................................................... E-19 4.3 Conclusions.......................................................................................................... E-21

5. ENVIRONMENTAL CONSIDERATION........................................................................... E-21

6. REFERENCES................................................................................................................. E-21 E-1

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR)

Paragraph 50.90, Southern Nuclear Operating Company (SNC) is requesting an amendment to the renewed licenses of Edwin I. Hatch Nuclear Plant (HNP), Units 1 and 2.

1

SUMMARY

DESCRIPTION The proposed amendment would modify the licensing basis to implement a change to the approved voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors." The proposed amendment would incorporate the use of an alternative seismic method, as described in this request, in addition to the peer reviewed, plant-specific HNP seismic probabilistic risk assessment (SPRA) into the previously approved 10 CFR 50.69 categorization process, as allowed by the Nuclear Regulatory Commission (NRC)-endorsed industry guidance.

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, evaluation and resolution of deviations). 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, existing treatment requirements are either maintained or potentially enhanced. Accordingly, 10 CFR 50.69 allows improved focus on equipment that has safety significance resulting in improved plant safety.

The 10 CFR 50.69 categorization process has been reviewed and approved by the NRC for Plant Hatch (Reference 4 ). Categorization includes an integrated assessment of total risk and the regulations and categorization guidance allows licensees to implement different approaches depending on the scope of their Probabilistic Risk Assessment (PRA) models. The currently approved risk assessment tools are:

1. Internal Event Risks using Internal Events and Internal Flooding PRA models

2. Fire Risks using Fire PRA models

3. Seismic Risks using Seismic PRA models

4. Other External Risks (e.g., tornados, external floods, etc.): An evaluation of external hazards was performed using Part 6 of the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard AS ME/ANS RA-Sa-2009 (Reference 1 ). These other external hazards were determined to be insignificant contributors to plant risk.

5. 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 2), which provides guidance for assessing and enhancing safety during shutdown operations.

This proposed amendment would apply the alternative seismic method using Electric Power Research Institute (EPRI) Report 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," dated February 2020 (Reference 3) as described in this request, in addition to the current use of a seismic PRA to assess seismic risk. The other aspects of the program remain as the NRC approved in Reference 4.

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change This request also includes an administrative change to delete a paragraph describing the actions to be completed prior to implementation of the 1 O CFR 50.69 License Amendment (Reference 4 ). As these one-time actions were completed in June 2020, prior to implementation of the 10 CFR 50.69 categorization process at HNP, this paragraph is no longer applicable and is proposed to be deleted from the revised license condition.

2 DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS The NRC has established a set of regulatory requirements for commercial nuclear reactors to enable a reactor facility to apply controls on structures, systems, and components (SSCs) that will provide reasonable assurance of adequate protection to public health and safety without imposing an undue risk to the health and safety of the public. 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 DB Es to protect public health and safety. Those 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. Special treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations. 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," is indirectly defined in 10 CFR 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants."

On November 22, 2004, a change to the NRC regulations was promulgated with the addition of 10 CFR 50.69. Section 50.69 represents an alternative set of requirements whereby a licensee may voluntarily undertake categorization of its SSCs, remove the listed special treatment requirements for SSCs that are determined to be of low individual safety significance, and implement specified alternative treatment requirements. The NRC's approval to use the 10 CFR 50.69 categorization process for Plant Hatch (Reference 4) included the addition of a new license condition to the Hatch RFOLs to allow implementation of 10 CFR 50.69, require prior NRC approval for a change to the categorization process, and specify implementation requirements that must be completed prior to implementation of the license amendment.

The previously approved HNP 50.69 categorization process conforms to the guidance in NRC Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1 dated May 2006 (Reference 5). The categorization process also conforms to the guidance in NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision O dated July 2005 (Reference 6), as endorsed by RG 1.201. With the proposed change to utilize the E-3

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change categorization process based upon the guidance in EPRI Report 3002017583 for Tier 2 facilities, and as described in this request, as an alternative to the Seismic PRA model, the HNP categorization process will continue to conform to these guidance documents.

2.2 REASON FOR PROPOSED CHANGE The SNC LAR (Reference 7) to allow HNP to implement the risk-informed categorization and treatment provisions of 10 CFR 50.69 identified the use of a peer reviewed plant-specific Seismic Probabilistic Risk Assessment (SPRA) model for the HNP categorization process for seismic hazards. The SNC risk management process provides assurance that the SPRA model reflects the as-built and as-operated Hatch plant in accordance with the PRA model maintenance process. However, SNC's experience has found that the maintenance of the SPRA model is more extensive than anticipated as the state of knowledge changes rapidly in the SPRA realm. As HNP only uses the SPRA in the 50.69 categorization process, the cost of maintaining the SPRA model for this sole application does not justify the benefits of its use. In addition, an alternative seismic method for EPRI Tier 2 sites has been made available to provide risk insights comparable to the use of SPRA, while employing the flexibility of a graded approach that supports the 10 CFR 50.69 categorization process.

The second paragraph of License Conditions 2.C.(11) and 2.C.(3)(i) of the HNP Unit 1 and Unit 2 Renewed Facility Operating Licenses, respectively, provide an allowance for a change to the categorization process, provided prior NRC approval, under 10 CFR 50.90, is obtained. Accordingly, this License Amendment Request is hereby submitted to obtain approval to implement the alternative seismic categorization process in addition to the SPRA method of SSC categorization for seismic hazards. When categorizing a system, Hatch will use a single approach (i.e., either SPRA or, if approved, the proposed alternate seismic approach using EPRI Report 3002017583 described in this request) for categorization of the entire system.

The HNP 10 CFR 50.69 categorization process has previously been reviewed and approved by NRC (Reference 4 ). The proposed change implements a modification to the process, as allowed by the 10 CFR 50.69 guidance endorsed by NRC in Regulatory Guide 1.201 (Reference 5), to use the categorization process based upon the guidance in EPRI Report 3002017583 for Tier 2 facilities and as described in this request (Reference 3).

2.3 DESCRIPTION

OF THE PROPOSED CHANGE SNC proposes the following changes to the license conditions for the operating licenses of HNP Unit 1 and Unit 2 that document the NRC's approval of the use 10 CFR 50.69 [i.e.,

Unit 1 Renewed Facility Operating License DPR-57 Condition 2.C.(11) and Unit 2 Renewed Facility Operating License NPF-5 Condition 2.C.(3)(i)], with bracketed text used to identify the unit-specific license amendment number(s) and the issuance date of the approved license amendment when it is issued.

Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. [305/250], dated June 26, 2020.

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change In addition, SNC is approved to implement 10 CFR 50.69 using the alternative seismic approach described in SNC's letter dated February 20, 2024, for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs, as specified in Renewed 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).

Prior to implementation of the Rene1Ned License Amendment No. [305/250], dated June 26, 2020, Southern Nuclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as built, as operated, and as maintained plant and shall ensure the risk acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met.

The second paragraph provides the requirements for the alternate seismic categorization process requested by this request. This text is underlined to identify it as new text. The proposed license condition markups are provided in the Attachment.

The implementation items identified in the last paragraph were completed as required by the original license condition prior to the implementation of the 1 O CFR 50.69 categorization process at HNP, which began in June 2020. Therefore, the paragraph specific to the implementation items is no longer applicable and is proposed to be deleted from the revised license condition.

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).

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change The above information was previously provided to NRC with the HNP application to allow the implementation of the provisions of 10 CFR 50.69 (Reference 7). The HNP 10 CFR 50.69 categorization process (overall process, including active and passive categorization elements) has been reviewed and approved by NRC (Reference 4).

In its review and approval of that application, NRC reviewed the technical adequacy of the HNP internal events (including internal flooding) at power, internal fire, and SPRA models and approved their use for 10 CFR 50.69 categorization (Reference 4 ). The HNP 50.69 process addresses seismic risk through the use of SPRA model, following the process defined in NEI 00-04 (Reference 6) and endorsed in RG 1.201 (Reference 5). The purpose of this license amendment request is to allow, within the approved HNP 50.69 program, the use of the alternative seismic method documented in the EPRI Report 3002017583 for Tier 2 sites and as described in this request in addition to the current use of a seismic PRA to assess seismic risk. Therefore, the remainder of this technical evaluation is focused on describing and evaluating the categorization process for this application using the alternative seismic method documented in EPRI Report 3002017583 for Tier 2 sites and as described in this request.

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

The alternative seismic categorization process as described in this request may be implemented for any system that was previously categorized and for systems that have not yet been categorized. However, any system that has been previously categorized is not required to be re-categorized with the alternate seismic categorization process as described in this request. When categorizing a system, Hatch will use a single approach (i.e., either SPRA or, if approved, the proposed alternative seismic method described in this request) for categorization of the entire system.

No assignments of Risk-Informed Safety Class (RISC) to SSCs are completed until a system is individually categorized, since categorization must be performed for entire systems and structures, not for selected components within a system or structure.

The categorization process using the alternative seismic method is discussed in the section 3.1.1 below.

3.1.1 Categorization Process Using Alternative Seismic Method for Tier 2 Sites 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 (such as Seismic Margin Analysis or Individual Plant Examination of External Events (IPEEE) screening) for determining SSC functional importance in the absence of a quantifiable PRA as part of an integrated, systematic process. For the HNP seismic hazard assessment, SNC 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 (Reference 6) sections 1.5 and 5.3. This approach is specified in EPRI report 3002017583 (Reference 3) with the EPRI markups provided in Attachment 2 of References 8 and 9 and includes additional considerations that are discussed in this section.

Note: The discussion below pertaining to Reference 3 includes the markups provided in of References 8 and 9.

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change EPRI Report 3002017583 (Reference 3) is an update to EPRI 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization,"

July 2018 (Reference 10) which was referenced in the NRC-issued amendment and safety evaluation (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 11 ).

(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 12), in particular, the response to the question regarding the differences between the initial EPRI report 3002012988 and the referenced EPRI Report 3002017583.

The proposed categorization approach for HNP 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 3 and plant specific insights considering seismic correlation effects and seismic interactions.

Following the criteria in Reference 3, the HNP site is considered a Tier 2 site because the site Ground Motion Response Spectrum (GMRS) to Design Basis Earthquake (DBE) comparison is above the Tier 1 threshold but not high enough that the NRC required the plant to perform a seismic PRA to respond to Recommendation 2.1 of the Near-Term Task Force 50.54(f) letter (Reference 28). As noted, Reference 3 also demonstrates that seismic risk is adequately addressed for Tier 2 sites by the results of additional qualitative assessments such as those 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 3, as amended by the request for additional information (RAI) responses and amendments (References 14, 15, 16, 17, 18, 19, 20, 21, and 22) 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 limited unique insights to the categorization process. That is the basis for the following statements in Table 4-1 of Reference 3.

"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 [high safety significant] 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 (IDP) for the final HSS determinations."

At sites with moderate seismic demands (i.e., Tier 2 range) such as HNP, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference 23. Tier 2 seismic demand sites have a lower likelihood of seismically induced failures and fewer challenges to plant systems than trial study plants. This lower likelihood is the technical basis for allowing use of a graded approach for addressing seismic hazards at HNP.

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change Test cases described in Section 3 of Reference 3, as amended by their RAI responses and amendments (References 14, 15, 16, 17, 18, 19, 20, 21, and 22) 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 3 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 3 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.

The test case information from Reference 3, developed by other licensees, including Case Study A (Reference 24 ), Case Study C (Reference 25), and Case Study D (Reference 26),

as well as RAI responses and amendments (References 14, 15, 16, 17, 18, 19, 20, 21, and 22) clarify aspects of these case studies and provide additional supporting bases for this application.

Basis for HNP Being a Tier 2 Plant As defined in Reference 3, HNP 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 3 applies to the Tier 2 sites in its entirety except for Sections 2.2 (Tier 1 sites) and 2.4 (Tier 3 sites).

While the Reference 3 guidance identifies the Tier 2 criterion based on a comparison of GMRS to SSE, for HNP, the DBE is the licensing basis earthquake that is used for this comparison. This substitution was acknowledged in Section 3.1, "Plant Seismic Design-Basis," of the NRC Staffs assessment of SNC information provided pursuant to the 10 CFR 50.54(f) request relating to Near-Term Task Force (NTTF) Recommendation 2.1 (Reference 27):

"For operating reactors licensed before 1997, the SSE is the plant licensing basis earthquake and is characterized by 1) a peak ground acceleration (PGA) value which anchors the response spectra at high frequencies (typically at 33 Hz) for the existing fleet of Nuclear Power Plants; (2) a response spectrum shape which depicts the amplified response at all frequencies below the PGA; and (3) a control point where the SSE is defined."

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 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 13).

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change The NRC issued its final determination of licensee seismic probabilistic risk assessments in a letter dated October 27, 2015 (Reference 28). The letter informed power reactor licensees of the remaining seismic evaluations to be performed and specifically informed those licensees that will perform an SPRA. In the letter, NRC stated:

"If the seismic hazard exceedance, peak of the spectral acceleration, and the general estimation of the SCDF were judged to be not significant, then the NRC staff concluded that a SPRA is not necessary for NRC's 50.54(f) letter-related regulatory decisions. Based on this additional assessment, the NRC staff has determined that SPRA are not warranted for 13 sites listed in Table 1 a in."

Note 3 of Table 1 a identifies HNP as a site that is not required to provide either an SPRA or a Seismic Margins Assessment (SMA).

As shown in Figure 1, comparing the HNP GMRS (derived from the seismic hazard) to the DBE (seismic design basis capability), the GMRS exceeds the DBE from 0.1 to 2 Hz and above 6 Hz. As such, it is appropriate that HNP is considered a Tier 2 plant. The basis for HNP being classified as Tier 2 will be documented and presented to the HNP IDP for each system that is categorized.

Hatch Response Spectra 1....----------,--,---------------------

0.6 -I---------'---'-------'-----.;........:.------

I-IP RLI:

GMRS Unit 2 DBE Uni l OBE o ~=~~~~~==

O.l 1

rn 100 Frequency [Hlz]

Figure 1: GMRS and DBE Response Spectra for Hatch Nuclear Plant, Units 1 & 2 (From Reference 29)

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

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change Implementation of the Recommended Process Reference 3 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 is the key element of the approach defined in Reference 3 for identifying unique seismic insights.

As discussed in Reference 51, 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 23) 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 provide margins well above the required design bases. As noted in Reference 51, experience has shown that design practices result in margins to realistic seismic capacities of 1.5 or more.

In applying the Reference 3 process for Tier 2 sites to the HNP 1 0 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the Reference 3 guidance and informed of plant SSC-specific seismic insights that the IDP may choose to consider in their HSS/low safety significance (LSS) deliberations. As part of the categorization team's preparation of the Risk-Based Document (RBD) 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 3 study and the bases for HNP 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 HNP 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 RBD. Integrated importance measures over the modeled hazards (i.e.,

internal events, including internal flooding, and internal fire for HNP) 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 HNP PRA models but having design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by E-10

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change seismic events, these will be addressed using non-PRA based qualitative assessments in conjunction with any seismic insights provided by the PRA.

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

The categorization team will review available HNP 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:

Impact of relay chatter Implications related to potential seismic interactions such as with block walls Seismic failures of passive SSCs such as tanks and heat exchangers Any known structural or anchorage issues with a particular SSC Components 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 3, including the markups provided in Attachment 2 of References 8 and 9, and described in this request. The process is summarized in Figure 2.

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change 2--

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r I

I Sa Identify SSCs I

I I

considered seismically 1----*1-----,

correlated I

I I

I Seismic Cap.icity and not subject t~ /

Sb Identify seismic interaction failures Yes-----...---,

I step 5 evaluations are per/armed I

I as part of a seismic walkdown

- No* -

J 7

Add seismic surrogate common cause groups to FPIE PRA T

S Quantify LOOP/SBO and S LOCA even~ and calculate FV and RAW Im porta nee Measures ssc.s Yes identified for correlation sscs meet FV or RAW thresholds for

/

~ :51/

Seismic Unique

.o'"t-----Yes_J HSSSSCs J,.

No con side ration Figure 2: Seismic Correlated Failure Assessment for Tier 2 Plants (Reproduced from Reference 3 (Figure 2-3), including the markups provided in of References 8 and 9)

Determination of seismic insights will make use of the FPIE 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:

E-12

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change Gather the population of SSCs in the system being categorized and review existing seismic information (Step 1 of Figure 2). This step may use the results of the required Tier 1 assessment that is performed along with the Tier 2 assessment. As stated in Reference 3 the technical basis for the Tier 1 approach in Section 2.2 of Reference 3 generally applies for Tier 2 plants in addition to the additional sensitivity and walkdowns described herein.

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

Perform a series of screenings to refine the list of SSCs subject to correlation sensitivity studies. Screens will identify (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 (Step 1 to Step Sb in Figure 2).

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 (Step 4 of Figure 2).

Perform Tier 2 Walkdown(s) focusing on identifying seismic correlated or interaction SSC failures for SSCs that were not previously walked down (Steps 5a/5b of Figure 2).

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 (Steps Sc/6 of Figure 2).

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 (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 3).

Quantify only the LOOP and small LOCA initiated accident sequences of the Tier 2 Adjusted PRA Model (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 (EOG) and DC power recovery).

E-13

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change 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 (Step 9 of Figure 2).

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

Prepare documentation of the Tier 2 analysis results, including identification of seismic unique HSS SSCs, for presentation to the IDP (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 3. 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 HNP 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 even if the HNP 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, SNC 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 Report 3002017583 SSC categorization criteria for the updated Tier. This includes use of the SNC corrective action process.

If the seismic hazard is reduced such that it meets the criteria for Tier 1 in EPRI Report 3002017583, SNC will implement the following process:

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

12.

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

If the seismic hazard increases to the degree that a SPRA becomes necessary to demonstrate adequate seismic safety, SNC will implement the following process using seismic PRA:

a. For previously completed system categorizations, SNC 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 E-14

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change categorization changes are warranted, they will be implemented through the SNC corrective action program 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 Report 3002017583 Section 2.4, "Tier 3 - High Seismic Hazard / Low Seismic Margin Sites."

Historical Seismic References for HNP The HNP GMRS and DBE curves from the seismic hazard and screening response are shown in Figure 1, which is replicated from the Expedited Seismic Evaluation Process (ESEP) report in Reference 29. The NRC's Staff review of the HNP ESEP report and responses is documented in Reference 42.

Section 1.1.3 of Reference 3 cites various post-Fukushima seismic reviews performed for the U.S. fleet of nuclear power plants. For HNP, the specific seismic reviews prepared by the licensee and the NRC's staff assessments are provided in the following licensee documents, which were submitted under oath and affirmation to the NRC.

1. Near-Term Task Force (NTTF) Recommendation 2.1 Seismic Hazard Screening (References 27, 28, 30, 31, and 50).

2. NTTF Recommendation 2.1 Spent Fuel Pool assessment (References 32 and 33).

3. NTTF Recommendation 2.3 Seismic Walkdowns (References 34, 35, 36, 37, 38, and 39).

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

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

(References 29, 40, 41, 42, and 43).

5. NTTF Recommendation 2.1 Seismic High Frequency Evaluation (References 44 and 45).

6. NTTF Recommendation 2.1 Seismic Low Frequency Evaluation (References 46 and 47)

Technical Information Precedent By letter dated January 31, 2020, Exelon Generation Company, LLC (EGC) submitted a license amendment request (Reference 52) to allow for the implementation of the provisions of 10 CFR 50.69 for LaSalle County Station (LSCS), Units 1 and 2. Following the criteria in EPRI report 3002012988 (Reference 10), based on the GM RS-to-SSE comparison, the LSCS site is considered a Tier 2 site, similar to the HNP site. For the LSCS seismic hazard assessment, EGC Nuclear also proposed the use of 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. EGC Nuclear provided responses to NRC requests for information (RAls) pertaining to the alternative seismic approach in letters dated October 1, 2020, October 16, 2020, and January 22, 2021 (References 49, 8, and 9, respectively). Reference 49 also included a response to an RAI that is not relevant to the alternative seismic approach and is not relevant to this HNP LAR. Based on information provided in Reference 52, as modified by the EGC Nuclear RAI responses in References E-15

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change 8, 9, and 49, NRC issued the license amendments approving the EGC request on May 27, 2021 (Reference 48).

SNC will follow the same alternative seismic approach in the 10 CFR 50.69 categorization process for HNP as that which was approved by the NRC staff for LaSalle County Station (Reference 48) with two clarifications:

This HNP LAR cites EPRI Report 3002017583 (Reference 3) with the markups provided in Attachment 2 of the LaSalle 10 CFR 50.69 LAR RAI responses dated October 16, 2020 (Reference 8) and January 22, 2021 (Reference 9) as applicable to the submittal. Additionally, the discussion above cites mark-ups to EPRI Report 3002012988 (Reference 10) that were submitted with References 8 and 9, whereas the proposed HNP alternative process is based on EPRI Report 3002017583 (Reference 3).

This HNP LAR also incorporates the LaSalle 1 O CFR 50.69 LAR RAI response (Reference 49) that addresses process issues associated with the proposed alternative seismic approach. However, the HNP LAR specifically excludes the response to LaSalle RAI APLC 50.69-RAI No. 12, because it addresses external events that are outside the scope of this request.

Summary Based on the above, the Summary from Section 2.3.3 of Reference 3 applies to HNP; namely, HNP 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 8, 9, and 49 (excluding the response to APLC 50.69-RAI No. 12, which addresses a non-seismic topic (external events)) 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 3 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 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii))

Section 3.2 of the SNC LAR to adopt the 10 CFR 50.69 provisions (Reference 7) demonstrated the adequacy of the quality and level of detail of the processes used to categorize SSCs. The technical adequacy of the HNP PRA was reviewed and confirmed to be acceptable for use in the SSC categorization process by the NRC in the License Amendment and associated Safety Evaluation allowing HNP Units 1 and 2 to implement the provisions of 10 CFR 50.69 (Reference 4 ). That technical adequacy evaluation is not impacted by the changes proposed in this request.

E-16

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))

Following submittal of the SNC LAR (Reference 7) to allow HNP to implement the risk-informed categorization and treatment provisions of 10 CFR 50.69, periodic model maintenance has been performed on the HNP PRA models to confirm these models continue to represent the as-built, as-operated plant. Maintenance of the internal flooding PRA model, which was completed in November 2018, resulted in a PRA Upgrade, triggering a Focused Scope Peer Review (FSPR). The FSPR for the internal flooding PRA model was performed in July 2019. The FSPR resulted into no Findings and one Suggestion Level Fact and Observation (F&O). In addition, the FSPR closed out two Finding Level F&Os for the internal flooding PRA that were issued when a full scope peer review was performed in November 2009.

3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv))

As described in Section 3.1.1, HNP may implement the alternative seismic categorization process. The processes identified in the current license condition may continue to be used.

The overall risk evaluation process described in NEI 00-04 (Reference 6) addresses both known degradation mechanisms and common cause interactions and meets the requirements of §50.69(b)(2)(iv). The sensitivity studies discussed in Section 8 of NEI 00-04, will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF).

The SSC failure rates and initiating event frequencies used in the HNP PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, human errors, etc.).

Subsequent performance monitoring and PRA updates as required by 10 CFR 50.69 will continue to include 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, a prompt 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.

To more specifically address the feedback and adjustment (i.e., performance monitoring) process as it pertains to the proposed alternative seismic method for Tier 2 sites discussed in Section 3.1.1 above, implementation of the SNC design control and corrective action programs provides assurance that the inputs for the qualitative determinations for seismic continue to remain valid to maintain compliance with the requirements of 10 CFR 50.69(e).

The performance monitoring process is described in SNC's 1 O CFR 50.69 program documents. The program requires that the periodic review assess changes that could E-17

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change impact the categorization results and provides the IDP with an opportunity to recommend categorization and treatment adjustments. Station personnel from engineering, operations, risk informed engineering, 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.

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

SNC has a comprehensive problem identification and corrective action program that requires the identification and resolution of issues. Any issue that may impact the 10 CFR 50.69 categorization process will be identified and addressed through the problem identification and corrective action program, including seismic-related issues.

The SNC 10 CFR 50.69 program requires that system categorization cannot be approved by the IDP until the panel's comments have been resolved to the satisfaction of the IDP.

This includes issues related to system-specific seismic insights considered by the IDP during categorization.

Scheduled periodic reviews no more frequent than once every two refueling outages will evaluate new insights resulting from available risk information (i.e., PRA model or other analysis used in the categorization) changes, design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated. This scheduled review will include:

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

In addition to the normally scheduled periodic reviews, if a PRA model or other risk information is upgraded, a review of the SSC categorization will be performed.

The periodic monitoring requirements of the 10 CFR 50.69 process will capture these issues and address them at a frequency commensurate with the issue severity. The 10 CFR 50.69 periodic monitoring program includes immediate and periodic reviews, that include the requirements of the regulation, to provide assurance that issues that could affect 10 CFR 50.69 categorization are addressed. The periodic monitoring process also monitors the performance and condition of categorized SSCs such that the assumptions for reliability in the categorization process are maintained.

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change 4

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

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

NRC Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance,"

Revision 1, May 2006.

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.

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

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

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS SNC proposes to modify the licensing basis to amend the approved voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors" to include use of an alternative seismic method in addition to the Hatch Nuclear Plant (HNP) Seismic Probabilistic Risk Assessment (SPRA).

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.

SNC 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 an alternative seismic method consistent with the U. S. Nuclear Regulatory Commission (NRC)-approved risk-informed categorization process to modify the scope of structures, systems, and components E-19

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change (SSCs) subject to the NRC regulatory treatment requirements and to implement alternative treatment requirements to be applied to SSCs. The alternative seismic method as described in this request is allowed by the 10 CFR 50.69 process guidance defined in the Nuclear Energy Institute (NEI) document NEI 00-04 as endorsed by the NRC in Regulatory Guide (RG) 1.201. The process used to evaluate SSCs for changes to NRC regulatory treatment requirements and the use of alternative requirements continue to provide assurance of the ability of the SSCs to perform their design functions. The potential change to treatment requirements does not change the design or 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 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 continues to permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC regulatory treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. The proposed change does not involve the installation of any additional plant equipment.

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 continue to permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC regulatory 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 a 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, SNC 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.

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change

4.3 CONCLUSION

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

5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6 REFERENCES

1.

ASME/ANS RA-Sa-2009, "Standard for Level I/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008," ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009

2.

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

3.

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]

4.

U. S. Nuclear Regulatory Commission to C.A. Gayheart (Southern Nuclear Operating Company), "Edwin I. Hatch Nuclear Plant, Units 1 and 2 - Issuance of Amendments Nos. 305 and 250, Regarding Adoption of 1 O CFR 50.69, 'Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors' (EPID L-2018-LLA-0175)," June 26, 2020 [ADAMS Accession No. ML20077J704]

5.

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

6.

NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, Nuclear Energy Institute, July 2005 [ADAMS Accession No. ML14365A203]

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change

7.

Gayheart, C.A. (Southern Nuclear Operating Company) to U. S. Nuclear Regulatory Commission, "Edwin I. Hatch Nuclear Plant, Units 1 & 2 - Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors'," June 7, 2018 [ADAMS Accession No. ML18158A583]

8.

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," October 16, 2020 [ADAMS Accession No. ML20290A791]

9.

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

10.

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

11.

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]

12.

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].

13.

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

March 12, 2012 [ADAMS Accession No. ML12053A340]

14.

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," August 28, 2018 [ADAMS Accession No. ML18240A065]

15.

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-001 O)," June 10, 2019

[ADAMS Accession No. ML19053A469]

16.

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]

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change

17.

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]

18.

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]

19.

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]

20.

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]

21.

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]

22.

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]

23.

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

24.

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]

25.

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 (RAls 4-11 )," February 21, 2018 [ADAMS Accession No. ML18052B342]

26.

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]

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Enclosure to NL-24-0051 Description and Assessment of the Proposed Change

27.

NRC Letter to Southern Nuclear Operating Company, Inc., "Edwin I. Hatch Nuclear Plant, Units 1 and 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident (TAC Nos. MF3772 and MF3773," April 27, 2015 [ADAMS Accession No. ML15097A424]

28.

NRC Letter to The Power Reactor Licensees on the Enclosed List, "Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 'Seismic' of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," October 27, 2015

[ADAMS Accession No. ML15194A015]

29.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.1," December 30, 2014 [ADAMS Accession No. ML15049A502]

30.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 Seismic Hazard and Screening Report for CEUS Sites,"

March 31, 2014 [ADAMS Accession No. ML14092A017]

31.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 Additional Information Regarding Seismic Hazard Curves,"

January 20, 2014 [ADAMS Accession No. ML15020A728]

32.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 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," December 15, 2016 [ADAMS Accession No. ML16350A350]

33.

NRC Letter to Southern Nuclear Operating Company, Inc., "Edwin I. Hatch Nuclear Plant - Units 1 and 2 - Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 (CAC Nos. MF3772 and MF 3773)," January 26, 2017

[ADAMS Accession No. ML17019A216]

34.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Unit 1 Seismic Recommendation 2.3 Walkdown Report 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 Daiichi Accident,"

March 12, 2012," November 27, 2012 [Package -ADAMS Accession No. ML123550172]

35.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Unit 2 Seismic Recommendation 2.3 Walkdown Report 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 Daiichi Accident, E-24

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change dated March 12, 2012," November 27, 2012 [Package -ADAMS Accession No. ML12355A636]

36.

NRC Letter to Those on the Enclosed List, "Request for Additional Information Associated with Near-Term Task Force Recommendation 2.3, Seismic Walkdowns,"

November 1, 2013 [ADAMS Accession No. ML 1330B418]

37.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to NRC Request for Additional Information Associated with Near-Term Task Force Recommendation 2.3, Seismic Walkdowns,"

November 25, 2016 [ADAMS Accession No. ML13330A556]

38.

NRC Letter to Southern Nuclear Operating Company, Inc., "Edwin I. Hatch, Unit 1 -

Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-lchi Nuclear Power Plant Accident (TAC No. MF0130)," June 6, 2014 [ADAMS Accession No. ML14155A361]

39.

NRC Letter to Southern Nuclear Operating Company, Inc., "Edwin I. Hatch, Unit 2 - Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-lchi Nuclear Power Plant Accident (TAC No. MF0131)," June 3, 2014

[ADAMS Accession No. ML14079A355]

40.

NRC Memorandum, "Edwin I. Hatch Nuclear Plant Units 1 and 2 - Technical Review Checklist Related to Interim Expedited Seismic Evaluation Process Supporting Implementation of NTTF Recommendation 2.1, Seismic, Related to the Fukushima Dai-lchi Nuclear Power Plant Accident (TAC No. MF5243 and MF5244)," July 7, 2015 [ADAMS Accession No. ML15190A131]

41.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 Request for Additional Information Regarding Expedited Seismic Evaluation Process Report," April 16, 2015 [ADAMS Accession No. ML15106A549]

42.

NRC Letter to Southern Nuclear Operating Company, Inc., "Edwin I. Hatch Nuclear Plant, Units 1 and 2 - Staff Review of Interim Evaluation Associated with Reevaluated Seismic Hazard Implementation of Near-Term Task Force Recommendation 2.1 (TAC Nos. MF5243 and MF5244)," July 22, 2015 [ADAMS Accession No. ML15201A474]

43.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 Fukushima Near-Term Task Force Recommendation 2.1 Expedited Seismic Evaluation Process Report Completion," December 15, 2016

[ADAMS Accession No. ML16350A329]

44.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope High Frequency Confirmation Evaluation," August 22, 2017

[ADAMS Accession No. ML17235B176]

45.

NRC Letter to Southern Nuclear Operating Company, Inc., "Edwin I. Hatch Nuclear Plant, Units 1 and 2 - Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," September 28, 2017 [ADAMS Accession No. ML17271A033]

E-25

Enclosure to NL-24-0051 Description and Assessment of the Proposed Change

46.

Southern Nuclear Operating Company, Inc. Letter to NRC, "Edwin I. Hatch Nuclear Plant - Units 1 and 2 Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope Low Frequency Evaluation," August 10, 2016 [ADAMS Accession No. ML16223A544]

47.

NRC Letter to Southern Nuclear Operating Company, Inc., "Edwin I. Hatch Nuclear Plant, Units 1 and 2 - Staff Review of Low Frequency Limited-Scope Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," October 20, 2016 [ADAMS Accession No. ML16285A421]

48.

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'," May 27, 2021 [ADAMS Accession No. ML21082A422)]

49.

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',"

October 1, 2020 [ADAMS Accession No. ML20275A292]

50.

"Hatch Unit 1 and 2 SPRA Relief, Information to Support Reconsideration of Hatch U1 for Inclusion in SPRA Relief due to Seismic Robustness and Similarities to Hatch U2," August 10, 2015 [ADAMS Accession No. ML15233A074]

51.

Exelon Generation Company, LLC Letter to NRC, "Limerick Generating Station, Units 1 and 2, Application to Implement an Alternate Defense-in-Depth Categorization Process, an Alternate Pressure Boundary Categorization Process, and an Alternate Seismic Tier 1 Categorization Process in Accordance with the Requirements of 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors'," March 11, 2021

[ADAMS Accession No. ML21070A412]

52.

Exelon Generation Company, LLC Letter to NRC, "LaSalle County Station, Units 1 and 2, Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power plants',"

January 31, 2020 [ADAMS Accession No. ML20031E699]

E-26

ATTACHMENT Edwin I. Hatch Nuclear Plant - Units 1 and 2 License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components Markups of HNP Units 1 & 2 Renewed Facility Operating Licenses A-1

Attachment to NL-24-0051 Markups of HNP Units 1 & 2 Renewed Facility Operating Licenses Edwin I. Hatch Nuclear Plant Unit 1, Renewed License No. DPR-57, License Condition 2.C.(11):

(11) 10 CFR 50.69 Risk-Informed Categorization Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. 305, dated June 26, 2020.

In addition, SNC is approved to implement 10 CFR 50.69 using the alternative seismic approach described in SNC's letter dated February 20, 2024, for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs, as specified in Renewed 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).

Prior to implementation of the Renewed License Amendment ~Jo. 305, dated June 26, 2020, Southern Nuclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as built, as operated, and as maintained plant and shall ensure the risl<: acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met Edwin I. Hatch Nuclear Plant Unit 2, Renewed License No. NPF-5, License Condition 2.C.{3){i):

(i) 10 CFR 50.69 Risk-Informed Categorization Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. 250 dated June 26, 2020.

In addition, SNC is approved to implement 10 CFR 50.69 using the alternative seismic approach described in SNC's letter dated February 20, 2024, for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs, as specified in Renewed License Amendment No. [YYY] 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).

Prior to implementation of the Rene1Ned License Amendment No. 250 dated June 26, 2020, Southern ~Juclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as built, as operated, and as maintained plant and shall ensure the risk acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met.

A-2