RA-23-0122, License Amendment Request to Revise the 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, Categorization Process to .

License Amendment Request to Revise the 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, Categorization Process to .

ML23229A456
Person / Time
Site:	Brunswick
Issue date:	08/17/2023
From:	Krakuszeski J Duke Energy Progress
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-23-0122
Download: ML23229A456 (1)

Text

John A. Krakuszeski Vice President Brunswick Nuclear Plant 8470 River Rd SE Southport, NC 28461 o: 910.832.3698 August 17, 2023 Serial: RA-23-0122 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324

Subject:

License Amendment Request to Revise the 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, Categorization Process to Reflect an Alternative Seismic Approach

References:

1. Letter from NRC to Duke Energy, Brunswick Steam Electric Plant, Units 1 and 2 -

Issuance of Amendment Nos. 292 and 320 to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2018-LLA-0008), dated September 17, 2019 (ADAMS Accession No. ML19149A471).

2. Letter from NRC to Duke Energy, Brunswick Steam Electric Plant, Units 1 and 2 -

Issuance of Amendment Nos. 305 and 333 to Revise License Conditions to Modify Approved 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors Categorization Process (EPID L-2020-LLA-0152), dated April 30, 2021 (ADAMS Accession No. ML21067A224).

Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy) hereby requests an amendment to the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2 Renewed Facility Operating Licenses (FOL). The proposed change would revise the license condition associated with the adoption 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, that was added to the BSEP FOL upon the issuance of Amendments 292 and 320 (Reference 1) and subsequently revised by Amendments 305 and 333 (Reference 2).

Specifically, the proposed change would revise the license condition to reflect an alternative approach to the one provided in Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, for evaluating the impact of the seismic hazard in the 10 CFR 50.69 categorization process.

The proposed alternative seismic approach for Tier 2 plants (e.g., BSEP) is described in Electric Power Research Institute (EPRI) 3002017583, Alternative Approaches for Addressing Seismic

U.S. Nuclear Regulatory Commission Page 2 Risk in 10 CFR 50.69 Risk-Informed Categorization," and is a risk-informed, graded approach that has demonstrated categorization insights equivalent to a seismic probabilistic risk assessment (PRA).

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

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed change involves no significant hazards consideration. The basis for this determination is provided in the Enclosure.

Duke Energy requests approval of the proposed amendment within one year of the date this submittal is accepted by the NRC staff for review. Once approved, Duke Energy will implement the license amendment within 120 days.

There are no regulatory commitments contained in this submittal.

In accordance with 10 CFR 50.91(b)(1), "Notice for Public Comment; State Consultation," a copy of this application, with the Enclosure and Attachment, 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 August 17, 2023.

Sincerely, John A. Krakuszeski JLV/jlv

Enclosure:

Description and Assessment of the Proposed Change

Attachment:

Facility Operating License Markup (Unit Nos. 1 and 2) cc:

Ms. Laura Dudes, Regional Administrator, Region II Mr. Luke Haeg, Project Manager Mr. Gale Smith, NRC Senior Resident Inspector Chair - North Carolina Utilities Commission Mr. Louis Brayboy, Radioactive Materials Branch Manager, N.C. DHHS

U.S. Nuclear Regulatory Commission Page 1 RA-23-0122 ENCLOSURE Description and Assessment of the Proposed Change

Subject:

License Amendment Request to Revise the 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, Categorization Process to Reflect an Alternative Seismic Approach

1.

SUMMARY

DESCRIPTION

2.

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

3.

TECHNICAL EVALUATION

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Analysis 4.4 Conclusions

5.

ENVIRONMENTAL CONSIDERATION

6.

REFERENCES ATTACHMENT: Facility Operating License Markup (Unit Nos. 1 and 2)

U.S. Nuclear Regulatory Commission Page 2 RA-23-0122

1.

SUMMARY

DESCRIPTION Duke Energy Progress, LLC (Duke Energy) hereby requests an amendment to the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2 Renewed Facility Operating Licenses (FOL).

The proposed change would revise the license condition associated with the adoption 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, that was added to the BSEP FOL upon the issuance of Amendments 292 and 320 (Reference 5) and subsequently revised upon the issuance of Amendments 305 and 333 (Reference 10).

Specifically, the proposed change would revise the license condition to reflect an alternative approach to the one provided in NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0 (Reference 6), for evaluating the impact of the seismic hazard in the 10 CFR 50.69 categorization process.

The proposed alternative seismic approach for Tier 2 plants (e.g., BSEP) is described in Electric Power Research Institute (EPRI) 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, (Reference 14) and is a risk-informed, graded approach that has demonstrated categorization insights equivalent to a seismic probabilistic risk assessment (PRA). For Tier 2 plants such as BSEP, the EPRI approach relies on the insights gained from the seismic PRAs examined in Reference 14 and plant specific insights considering seismic correlation effects and seismic interactions.

Furthermore, in Attachment 1 of Reference 26, various implementation items were identified that required completion prior to the implementation of 10 CFR 50.69 at BSEP. Duke Energy has completed those implementation items.

2.

DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS In Reference 5, the NRC issued Amendments 292 and 320 to the FOL for BSEP, which added a new license condition to allow for the implementation of the provisions of 10 CFR 50.69. The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems and components (SSCs) subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment and evaluation) based on a method of categorizing SSCs according to their safety significance.

Consistent with the guidance in NEI 00-04, as endorsed by Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants according to their Safety Significance, (Reference 7) the existing BSEP categorization process uses PRAs to assess risk from internal events (including internal flooding), fire, and high winds. In Reference 10, the NRC issued Amendments 305 and 333 to revise the 10 CFR 50.69 license condition to rely on external flood hazard screening in the 10 CFR 50.69 categorization process instead of utilizing the external flood PRA. BSEP currently uses the Seismic Safe Shutdown Equipment List (SSEL) from the seismic margin analysis (SMA) to assess seismic risk, the individual plant examination of external events (IPEEE) screening process to assess other external hazards, and a qualitative defense-in-depth shutdown model to assess shutdown risk.

Regarding the subject license amendment request and the assessment of seismic risk in the 10 CFR 50.69 categorization process, BSEP currently uses the SMA screening method. BSEP

U.S. Nuclear Regulatory Commission Page 3 RA-23-0122 currently follows the approach in Reference 6 using the SSEL to identify credited equipment as high safety significance (HSS), regardless of the equipments capacity, frequency of challenge or level of functional diversity. Consistent with Reference 6, the BSEP 10 CFR 50.69 categorization process considers all components in the SSEL as HSS based on seismic risk.

All components not listed in the SSEL are considered preliminary low safety significant (LSS) with respect to seismic risk.

2.2 REASON FOR PROPOSED CHANGE Reference 7 clarifies that the NRC staff expects that licensees proposing to use non-PRA approaches in the 10 CFR 50.69 categorization process provide a basis in the submittal for why the approach and the accompanying method employed to assign safety significance to SSCs is technically adequate. The guidance further states that as part of the NRCs review and approval of an application requesting to implement 10 CFR 50.69, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA methods used in the categorization approach. To that end, the NRC imposed a license condition on BSEP in Reference 5 that requires use of the SMA to evaluate seismic risk in the 10 CFR 50.69 categorization process.

Since Duke Energy desires to change the categorization approach specified in the BSEP FOL with respect to the assessment of seismic risk (i.e., switch from a SMA to the alternative EPRI approach in Reference 14), NRC approval of the new approach in the categorization process must be requested pursuant to 10 CFR 50.90, in accordance with the license condition that was added upon issuance of BSEP Amendment Nos. 292 and 320, and carried forward upon the issuance of Amendment Nos. 305 and 333.

This license amendment request follows the same categorization approach for Tier 2 seismic risk as approved for LaSalle County Station, Units 1 and 2 (Reference 15) and Fitzpatrick Nuclear Plant (Reference 61). Deviations from the approach taken by those licensees are limited to details in Duke Energy's configuration control associated with the 10 CFR 50.69 categorization process, which is further discussed in Section 3.5 of this enclosure.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE Duke Energy proposes to revise the 10 CFR 50.69 license condition that was added to the BSEP Unit 1 FOL by Amendment 292 (Reference 5) and revised upon the issuance of Amendment 305 (Reference 10) to the following:

Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 License Amendment No. 305 dated April 30, 2021.

U.S. Nuclear Regulatory Commission Page 4 RA-23-0122 In addition, Duke Energy is approved to implement 10 CFR 50.69 using the alternative seismic approach for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs as described in Duke Energy letter RA-23-0122, dated August 17, 2023, as specified in Unit 1 License Amendment No. [XXX] dated [DATE].

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

Duke Energy proposes to revise the Implementation Date associated with the license condition to the following:

Upon implementation of Amendment No. [XXX]

Duke Energy proposes to revise the 10 CFR 50.69 license condition that was added to the BSEP Unit 2 FOL by Amendment 320 (Reference 5) and revised upon the issuance of Amendment 333 (Reference 10) to the following:

Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 2 License Amendment No. 333 dated April 30, 2021.

In addition, Duke Energy is approved to implement 10 CFR 50.69 using the alternative seismic approach for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs as described in Duke Energy letter RA-23-0122, dated August 17, 2023, as specified in Unit 2 License Amendment No. [XXX] dated [DATE].

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

Duke Energy proposes to revise the Implementation Date associated with the license condition to the following:

Upon implementation of Amendment No. [XXX]

A markup of the BSEP Unit Nos. 1 and 2 FOL to describe the proposed change is provided in the Attachment.

The proposed FOL language would credit use of both the SMA and alternate Tier 2 methods consistent with the approved license condition for Shearon Harris Nuclear Power Plant, Unit 1 (Reference 9) that allows both the SMA and Tier 1 methods to be implemented at the site.

U.S. Nuclear Regulatory Commission Page 5 RA-23-0122 Before providing the technical evaluation of the alternate seismic approach proposed to be used in the BSEP Unit Nos. 1 and 2 10 CFR 50.69 categorization process, it is necessary to disposition the PRA implementation items that are referenced in the existing BSEP Unit Nos. 1 and 2 10 CFR 50.69 license conditions. The existing license conditions that were added with the issuance of Amendments 305 and 333 to the BSEP Units 1 and 2 FOL states the following, in part:

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to the NRC dated November 24, 2020, prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above. The issues identified in the attachment will be addressed and any associated changes will be made prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

Duke Energy confirms that implementation items i. through vi. listed in the November 24, 2020 Duke Energy letter (Reference 26), as described in the NRC Safety Evaluation (SE) approving 10 CFR 50.69 for BSEP Unit Nos. 1 and 2 (Reference 10), are complete.

Thus, the proposed change removes the discussion of the Reference 26 implementation items from the license conditions for BSEP Unit Nos. 1 and 2.

Duke Energy also confirms that all other previously approved 10 CFR 50.69 categorization methods for BSEP in Reference 5 and Reference 10, except for the proposed adoption of an alternate seismic approach, are not impacted by this license amendment request.

3.

TECHNICAL EVALUATION For the proposed change, the process to categorize each system will continue to be consistent with the guidance in NEI 00-04, as endorsed by RG 1.201, with the exception of the evaluation of impact of the seismic hazard, which will use the EPRI 3002017583 (Reference 14) approach for seismic Tier 2 sites to assess seismic hazard risk for 10 CR 50.69. Inclusion of additional process steps to address seismic considerations, as discussed below, will ensure that reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(iv) is achieved.

Duke Energy proposes that the alternate Seismic Tier 2 categorization process may be implemented for any BSEP system that was previously categorized or for systems that will be categorized. However, any system that has been previously categorized is not required to be re-categorized with the alternate Seismic Tier 2 categorization process. The processes identified in the existing NRC-approved 10 CFR 50.69 license condition may continue to be used. With the proposed change, BSEP will use a single approach for a given system categorization (e.g., either SMA or alternate seismic approach described herein).

Table 1 below is an update to Table 3-1 in BSEP's original submittal to adopt 10 CFR 50.69 (Reference 1). Changes are marked by bold, italic font to highlight the proposed alternative approach for evaluating the seismic hazard.

U.S. Nuclear Regulatory Commission Page 6 RA-23-0122 Table 1: Categorization Evaluation Summary Element Categorization Step -

NEI 00-04 Section Evaluation Level IDP Change HSS to LSS Drives Associated Functions Risk (PRA Modeled)

Internal Events Base Case - Section 5.1 Component Not Allowed Yes Fire, Seismic and Other External Events Base Case Allowable No PRA Sensitivity Studies Allowable No Integral PRA Assessment - Section 5.6 Not Allowed Yes Risk (Non-modeled)

Fire, and Other External Hazards -

Component Not Allowed No Seismic -

SMA Process Component Not Allowed No Seismic -

Alternative Tier 2 Approach Function/Component Allowed1 No Shutdown - Section 5.5 Function/Component Not Allowed No Defense-in-Depth Core Damage -

Section 6.1 Function/Component Not Allowed Yes Containment - Section 6.2 Component Not Allowed Yes Qualitative Criteria Considerations -

Section 9.2 Function Allowable N/A Passive Passive - Section 4 Segment/Component Not Allowed No 1 Integrated Decision-making Panel (IDP) consideration of seismic insights can also result in an LSS to HSS determination.

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 SMA or IPEEE Screening) as part of an integrated, systematic process. For the BSEP 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

U.S. Nuclear Regulatory Commission Page 7 RA-23-0122 alternative to those listed in NEI 00-04 sections 1.5 and 5.3. This approach is specified in Reference 14 with the EPRI markups provided in Attachment 2 of References 23 and 24 and includes additional considerations that are discussed in this section.

Note: The discussion below pertaining to Reference 14 includes the markups provided in Attachment 2 of References 23 and 24.

EPRI 3002017583 (Reference 14) is an update to EPRI 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," July 2018 (Reference 62) 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:

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," February 28, 2020 (Reference 20).

This BSEP license amendment incorporates 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 63), 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 BSEP 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 14 and plant specific insights considering seismic correlation effects and seismic interactions.

Following the criteria in Reference 14, BSEP is considered a Tier 2 site because the site Ground Motion Response Spectrum (GMRS) to SSE (Safe Shutdown Earthquake) 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 (NTTF) 50.54(f) letter (Reference 28). Reference 14 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 14, as amended by their request for additional information (RAI) responses and NRC issued amendments (References 46, 47, 48, 49, 50, 51, 52, 53, and 54) demonstrate 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 14.

"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 PRA can identify

U.S. Nuclear Regulatory Commission Page 8 RA-23-0122 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 BSEP, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference 29. 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 BSEP.

Test cases described in Section 3 of Reference 14, as amended by their RAI responses and NRC issued amendments (References 46, 47, 48, 49, 50, 51, 52, 53, and 54) demonstrated 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 14 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 14 uses common cause failures, similar to the approach taken in a full power internal events (FPIE) PRA and can identify the appropriate seismic insights to be considered by the IDP along with the other categorization insights for the final HSS determinations.

Duke Energy is using test case information from Reference 14, developed by other licensees.

The test case information is being incorporated by reference into this application, specifically Case Study A (Reference 55), Case Study C (Reference 56), and Case Study D (Reference 57). This includes RAI responses and NRC issued amendments (References 46, 47, 48, 49, 50, 51, 52, 53, and 54), that clarify aspects of these case studies.

Basis for BSEP being a Tier 2 Plant As defined in Reference 14, BSEP 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 14 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 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 seismic PRA to respond to the Fukushima 10 CFR 50.54(f) letter (Reference 28).

As shown in Figure 1, comparing the BSEP GMRS (derived from the seismic hazard) to the SSE (i.e., seismic design basis capability), in a portion of the 1 to 10 Hz part of the response spectrum, the GMRS exceeds the SSE. The GMRS is exceeded by the IHS (IPEEE HCLPF

U.S. Nuclear Regulatory Commission Page 9 RA-23-0122

[High-Confidence-of-Low-Probability-of-Failure] Spectrum) in the 1 to 10 Hz part of the response spectrum. For a portion of the range above 10 Hz, the GMRS exceeds the IHS.

Since the BSEP SSE is exceeded by the updated GMRS within the 1 to 10 Hz range (Figure 1),

further seismic evaluation was conducted and documented within the Reference 30 submittal to the NRC to demonstrate adequacy in BSEP's seismic evaluation such that a full seismic PRA was not required. This screening was supported by incorporating consideration of the IHS comparison to the GMRS. The NRC screened out BSEP from performing an seismic PRA in response to the NTTF 2.1 10 CFR 50.54(f) letter, as documented in Reference 31.

Additionally, a High Frequency Confirmation was performed (Reference 32) which reported:

The supplemental calculation for BSEP shows that the GMRS exceedance area between the GMRS and IHS/Review Level Earthquake (RLE) is on the order of 10 percent or less of the area under the IHS/RLE over the frequency range of exceedance.

As such, the GMRS exceedance is consistent with the criteria identified in Section 3.1.2 of Reference 33, and no additional evaluation is necessary.

As such, the Brunswick seismic hazard meets the criteria for Tier 2 from Reference 14. The basis for BSEP being a Tier 2 site will be documented and presented to the IDP for each system categorized.

Figure 1: Comparison of BSEP GMRS to the Safe Shutdown Earthquake (SSE), the Review Level Earthquake (RLE) at the Surface, and the IPEEE HCLPF Spectrum (IHS)

U.S. Nuclear Regulatory Commission Page 10 RA-23-0122 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.

Implementation of the Recommended Process Reference 14 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 14 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 29) 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 14 process for Tier 2 sites to the BSEP 10 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the Reference 14 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 14 study and the bases for BSEP 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 BSEP 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 the modeled hazards (i.e., internal events, including internal flooding, internal fire, and high winds for BSEP) 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

U.S. Nuclear Regulatory Commission Page 11 RA-23-0122 BSEP PRA models but having design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events, these will be addressed using non-PRA based qualitative assessments in conjunction with any seismic insights provided by the PRA.

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

The categorization team will review available BSEP plant-specific seismic insights 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 14 and is summarized below in Figure 2.

U.S. Nuclear Regulatory Commission Page 12 RA-23-0122 Figure 2: Seismic Correlated Failure Assessment for Tier 2 Plants Note - Reproduced from Reference 14: Figure 2-4.

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:

o 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 required Tier 1 assessment that is performed along with the Tier 2 assessment. As stated in Reference 14 the technical basis for the Tier 1 approach in Section 2.2 of

U.S. Nuclear Regulatory Commission Page 13 RA-23-0122 Reference 14 generally applies for Tier 2 plants in addition to the additional sensitivity and walkdowns described herein.

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

o 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):

Inherently rugged SSCs SSCs not in Level 1 or Level 2 PRAs Components already identified as HSS components from the Internal Events PRA or Integrated assessment o 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).

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

o 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).

o 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 14).

o 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).

o 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).

U.S. Nuclear Regulatory Commission Page 14 RA-23-0122 o SSCs screened out in Steps 5c, 6, or 9 in Figure 2 can be considered LSS (reference Step 10 of Figure 2).

o 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 14. 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 with a means to consider potential impacts of seismic events in the categorization process.

If the BSEP seismic hazard changes from medium risk (i.e., Tier 2) at some future time and the 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, prior NRC approval, pursuant to 10 CFR 50.90, will be requested. Upon receipt of NRC approval for such a change, 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 process, 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 3002017583 Section 2.2, "Tier 1 - Low Seismic Hazard / High Seismic Margin Sites."

If the seismic hazard increases to the degree that a seismic PRA becomes necessary to demonstrate adequate seismic safety, Duke Energy will implement the following process following completion of the seismic PRA, including adequate closure of Peer Review Findings and Observations:

a) For previously completed system categorizations, Duke Energy will review the categorization results using the seismic PRA 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 process, corrective action program, and NEI 00-04 Section 12.

U.S. Nuclear Regulatory Commission Page 15 RA-23-0122 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 BSEP The BSEP 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 30). The NRC's Staff assessment of the BSEP seismic hazard and screening response is documented in Reference 34. In the Staff Confirmatory Analysis (Section 3.3.3) of Reference 34, the NRC concluded that the methodology used by Duke Energy adequately characterizes the seismic hazard for the BSEP site.

Section 1.1.3 of Reference 14 cites various post-Fukushima seismic reviews performed for the U.S. fleet of nuclear power plants. For BSEP, the specific seismic reviews prepared by Duke Energy and the NRC's staff assessments of those reviews are provided here. These licensee documents were submitted under oath and affirmation to the NRC.

1. NTTF Recommendation 2.1 Seismic Hazard Screening (References 35 and 34).

2. NTTF Recommendation 2.1 Spent Fuel Pool assessment (References 36 and 37).

3. NTTF Recommendation 2.3 Seismic Walkdowns (References 38, 39, and 40).

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

(References 41 and 42).

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

5. NTTF Recommendation 2.1 Expedited Seismic Evaluation Process (ESEP) (References 43 and 44).

6. NTTF Recommendation 2.1 Seismic High Frequency Evaluation (References 32 and 45).

Summary Based on the above, the Summary from Section 2.3.3 of Reference 14 applies to BSEP; namely, BSEP 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 23, 24, and 22 are incorporated by reference into this amendment request as they provide additional supporting bases for Tier 2 plants such as BSEP to adopt the alternative seismic methodology for use in the 10 CFR 50.69 categorization process. The Tier 2 alternate seismic methodology was previously approved for LaSalle County Station, Units 1 and 2 (LaSalle) in Reference 15.

Similarly, References 59 and 60 for James A. FitzPatrick Nuclear Power Plant (FitzPatrick) are incorporated by reference into this amendment request as they also provide supporting bases for Tier 2 plants such as BSEP to adopt the alternative seismic methodology for use in the 10

U.S. Nuclear Regulatory Commission Page 16 RA-23-0122 CFR 50.69 categorization process. The Tier 2 alternate seismic methodology was previously approved for Fitzpatrick in Reference 61.

In addition, References 18, 19, 20, and 58 are incorporated by reference into this amendment request as they provide additional supporting bases for Tier 1 plants that are also used for Tier 2 plants. The special sensitivity study recommended using common cause failures, similar to the approach taken in a FPIE PRA, which 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 14 to assess seismic hazard risk for 10 CFR 50.69, along 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).

Status of the BSEP PRA Models Since the original BSEP 50.69 LAR (Reference 1) was approved by the NRC staff (Reference 5), model updates and finding closures have occurred. The related model changes were limited to PRA maintenance activities.

The finding level Facts and Observations (F&Os) related to the BSEP internal events and internal flooding PRA models that were captured in Attachment 3 of the BSEP 10 CFR 50.69 LAR submittal (Reference 1) were reviewed and closed in December 2019 and May 2020 using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations, as accepted by the NRC in the letter dated May 3, 2017 (Reference 27). No PRA upgrades were identified by the F&O Closure review teams. After all the reviews, all Supporting Requirements (SRs) for the BSEP internal events and internal flooding PRA models were assessed to be met at least at Capability Category II and there are no open finding level F&Os.

The finding level F&Os related to the BSEP fire PRA model that were captured in Attachment 3 of the BSEP 10 CFR 50.69 LAR submittal (Reference 1) were reviewed and closed in August 2018 using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, as accepted by the NRC in the letter dated May 3, 2017 (Reference 27). No PRA upgrades were identified by the F&O Closure review teams. After all the reviews, all SRs for the BSEP fire PRA model were assessed to be met at least at Capability Category II and there are no open finding level F&Os.

Consistent with the original BSEP 10 CFR 50.69 LAR (Reference 1) submittal, there are no open findings for the BSEP High Winds PRA model.

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

U.S. Nuclear Regulatory Commission Page 17 RA-23-0122 The Duke Energy configuration control process ensures that changes to the plant, including a physical change to the plant and changes to documents, are evaluated to determine the impact to drawings, design bases, licensing documents, programs, procedures, and training.

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 categorization process will be identified and addressed through the problem identification and corrective action program.

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

All other aspects of the BSEP 10 CFR 50.69 Feedback and Review process remain as stated in the original BSEP 10 CFR 50.69 LAR (Reference 1) which was reviewed and approved by the NRC staff (Reference 5).

The Periodic Review process, as described in the original BSEP 10 CFR 50.69 LAR (Reference

1) and RAI responses (Reference 2), assesses system/component performance changes and plant operation or design changes that have occurred for categorized systems on a frequency no longer than once every two refueling outages, as required by 10 CFR 50.69(e), to review the impact of plant changes on RISC-1, RISC-2, RISC-3, and RISC-4 SSCs. The review is implemented by Duke Energy fleet procedure.

Technical Information Incorporated by Reference For BSEP, Duke Energy proposes to apply the same alternate seismic Tier 2 methodologies in the 10 CFR 50.69 categorization process that was approved by the NRC staff for LaSalle (Reference 15), with the following exceptions:

1) The site-specific LaSalle information (e.g., seismic capacity discussions, etc.) from the 10 CFR 50.69 and other LaSalle licensing responses do not apply to BSEP. BSEP site-specific seismic capacity information is described above herein.

2) The configuration control checklist described in the LaSalle submittal implies that a specific checklist was developed for 10 CFR 50.69 reviews. Refer to the discussion in the previous section of this submittal for the configuration control and periodic review processes that are employed by Duke Energy for BSEP.

4.

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

The regulations in 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

U.S. Nuclear Regulatory Commission Page 18 RA-23-0122 NRC Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, May 2006. (Reference 7)

NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, January 2018. (Reference 12)

NRC Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020. (Reference 13)

The alternate seismic Tier 2 categorization process is consistent with the applicable regulations in 10 CFR 50.69. The proposed change represents a deviation from the NEI 00-04 guidance endorsed in RG 1.201, Revision 1. However, the NRC staff specifies that licensees may propose alternative approaches for the 10 CFR 50.69 categorization process and should provide a basis in the submittal explaining why the approach and the accompanying method employed to assign safety significance to SSCs is technically acceptable. The intent of the technical evaluation provided above in Section 3 of the subject LAR is to satisfy the NRC expectation cited in RG 1.201, Revision 1 regarding the proposed alternate seismic Tier 2 approach to be used in the BSEP 10 CFR 50.69 categorization process.

4.2 PRECEDENT The NRC previously issued license amendments to LaSalle and FitzPatrick to adopt the same alternate seismic Tier 2 methodologies in the 10 CFR 50.69 categorization process that are proposed in this BSEP LAR.

LaSalle County Station, Units 1 and 2: Application dated January 31, 2020 (ADAMS Accession No. ML20031E699) with relevant RAI responses dated October 1, 2020 (ADAMS Accession No. ML20275A292), October 16, 2020 (ADAMS Accession No. ML20290A791), and January 22, 2021 (ADAMS Accession No. ML21022A130); NRC Safety Evaluation dated May 27, 2021 (ADAMS Accession No. ML2108A422).

James A. Fitzpatrick Nuclear Power Plant: Application dated July 30, 2021 (ADAMS Accession No. ML21211A078) with supplement dated March 4, 2022 (ADAMS Accession No. ML22063A135); NRC Safety Evaluation dated August 23, 2022 (ADAMS Accession No. ML22196A061).

4.3 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS Duke Energy Progress, LLC (Duke Energy) requests an amendment to the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2 Renewed Facility Operating Licenses (FOL). The proposed change would revise the license condition associated with the adoption 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, that was added to the BSEP FOL upon the issuance of Amendments 292 and 320 and revised upon the issuance of Amendments 305 and 333. Specifically, the proposed change would revise the license condition for each unit to reflect an alternative approach to the one provided in NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, to allow the use of an alternate Seismic Tier 2 categorization process in the 10 CFR 50.69 categorization process.

U.S. Nuclear Regulatory Commission Page 19 RA-23-0122 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 would revise the BSEP license condition for each unit that was added with the issuance of BSEP license Amendments 292 and 320 and revised upon the issuance of Amendments 305 and 333 to reflect use of an alternate Seismic Tier 2 categorization process in the 10 CFR 50.69 categorization process. The approach is described in EPRI 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization. With the proposed change, BSEP will continue to be permitted to use 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 will continue to ensure the ability of the SSCs to perform their design function. The potential change to special treatment requirements using the alternate seismic Tier 2 categorization process does not change the design and operation of the SSCs. As a result, the proposed change to revise the 10 CFR 50.69 categorization process to reflect an alternate Seismic Tier 2 methodology 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 would revise the BSEP license condition for each unit that was added with the issuance of BSEP license Amendments 292 and 320 and revised upon the issuance of Amendments 305 and 333 to reflect use of an alternate Seismic Tier 2 categorization process in the 10 CFR 50.69 categorization process. The approach is described in EPRI 3002017583. With the proposed change, BSEP will continue to be permitted to use a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

U.S. Nuclear Regulatory Commission Page 20 RA-23-0122 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 would revise the BSEP license condition for each unit that was added with the issuance of BSEP license Amendments 292 and 320 and revised upon the issuance of Amendments 305 and 333 to reflect use of an alternate Seismic Tier 2 categorization process in the 10 CFR 50.69 categorization process. The approach is described in EPRI 3002017583. With the proposed change, BSEP will continue to be permitted to use 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.

4.4 CONCLUSION

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

5.

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

U.S. Nuclear Regulatory Commission Page 21 RA-23-0122

6.

REFERENCES

1. Duke Energy letter to NRC, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, January 10, 2018 (ADAMS Accession No. ML18010A344).

2. Duke Energy letter to NRC, Response to NRC Request for Additional Information (RAI)

Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, November 2, 2018 (ADAMS Accession No. ML18306A523).

3. Duke Energy letter to NRC, Response to NRC Request for Additional Information (RAI)

Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, February 13, 2019 (ADAMS Accession No. ML19044A366).

4. Duke Energy letter to NRC, Response to NRC Request for Additional Information (RAI)

Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, April 8, 2019 (ADAMS Accession No. ML19099A035).

5. NRC letter to Duke Energy, Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment Nos. 292 and 320 to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2018-LLA-0008), September 17, 2019 (ADAMS Accession No. ML19149A471).

6. NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, Nuclear Energy Institute, July 2005 (ADAMS Accession No. ML052910035).

7. 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.

8. [Not Used]

9. NRC letter to Duke Energy, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment No. 188 Regarding Revision of the 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, Categorization Process to Reflect an Alternative Seismic Approach (EPID L-2021-LLA-0003), January 19, 2022 (ADAMS Accession No. ML21316A248).

10. NRC letter to Duke Energy, Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment Nos. 305 and 333 to Revise License Conditions to Modify Approved 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors Categorization Process (EPID L-2020-LLA-0152), April 30, 2021 (ADAMS Accession No. ML21067A224).

U.S. Nuclear Regulatory Commission Page 22 RA-23-0122

11. 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).

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

13. NRC Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020.

14. 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).

15. NRC letter to Exelon Generation Company, 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 (EPID L-2020-LLA-0017)," May 27, 2021 (ADAMS Accession No. ML21082A422).

16. Exelon Generation Company (Calvert Cliffs) letter to NRC, Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors, November 28, 2018 (ADAMS Accession No. ML18333A022).

17. Exelon Generation Company (Calvert Cliffs) letter to NRC, Revised submittal to Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors, May 10, 2019 (ADAMS Accession No. ML19130A180).

18. Exelon Generation Company (Calvert Cliffs) letter to NRC, Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors, July 1, 2019 (ADAMS Accession No. ML19183A012).

19. Exelon Generation Company (Calvert Cliffs) letter to NRC, Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors, July 19, 2019 (ADAMS Accession No. ML19200A216).

20. NRC letter to Exelon Generation Company, 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).

U.S. Nuclear Regulatory Commission Page 23 RA-23-0122

21. Exelon Generation Company (LaSalle) letter to NRC, Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors, January 31, 2020 (ADAMS Accession No. ML20031E699).

22. Exelon Generation Company 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 (EPID L-2020-LLA-0017), October 1, 2020 (ADAMS Accession No. ML20275A292).

23. Exelon Generation Company 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 (EPID L-2020-LLA-0017), October 16, 2020 (ADAMS Accession No. ML20290A791).

24. Exelon Generation Company (LaSalle) letter to NRC, Response to Request for Additional Information Regarding the License Amendment Request to Adopt 10 CFR 50.69 (EPID L-2020-LLA-0017), January 22, 2021 (ADAMS Accession No. ML21022A130).

25. [Not Used]

26. Duke Energy letter to NRC, Response to Request for Additional Information (RAI) for License Amendment Request to Modify Approved 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors Categorization Process, November 24, 2020 (ADAMS Accession No. ML20329A466).

27. Letter from NRC to 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).

28. 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).

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

30. Duke Energy letter to NRC, Seismic Hazard and Screening Report (CEUS Sites),

Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 31, 2014. (ADAMS Accession No. ML14106A461).

U.S. Nuclear Regulatory Commission Page 24 RA-23-0122

31. NRC Letter to Duke Energy, "Brunswick Steam Electric Plant, Units 1 and 2 - Screening and Prioritization Results of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (TAC Nos. MF3824 and MF3825)," September 17, 2014 (ADAMS Accession No. ML14231A964).

32. Duke Energy letter to NRC, High Frequency Supplement to Seismic Hazard Screening 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. ML16365A024).

33. Electric Power Research Institute (EPRI) Technical Report 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility Evaluation, July 2015 (ADAMS Accession No. ML15223A102).

34. NRC Letter to Duke Energy, "Brunswick Steam Electric Plant, Units 1 and 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (CAC Nos. MF3824 and MF3825)," March 1, 2016 (ADAMS Accession No. ML16041A435).

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

Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," March 31, 2014 (ADAMS Accession No. ML14106A461).

36. 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," December 15, 2016 (ADAMS Accession No. ML16365A025).

37. NRC Letter to Duke Energy, "Brunswick Steam Electric 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. MF3824 and MF3825)," February 2, 2017 (ADAMS Accession No. ML17031A001).

38. Duke Energy Letter to NRC, "Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," November 27, 2012 (ADAMS Accession No. ML12349A388).

39. NRC letter to Duke Energy, "Brunswick Steam Electric Plant, Unit 2 - Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident (TAC No. MF0100)," March 4, 2014 (ADAMS Accession No. ML14050A031).

U.S. Nuclear Regulatory Commission Page 25 RA-23-0122

40. NRC letter to Duke Energy, "Brunswick Steam Electric Plant, 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 (TAC No. MF0099)," March 4, 2014 (ADAMS Accession No. ML14050A030).

41. Duke Energy letter to NRC, 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, August 17, 2017. (ADAMS Accession No. ML17229B504).

42. NRC letter to Duke Energy, "Brunswick Steam Electric Plant, Units 1 and 2 - Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC Nos. MF7807 and MF7808; EPID L-2016-JLD-0006)," January 30, 2018 (ADAMS Accession No. ML18017A121).

43. Duke Energy letter to NRC, Expedited Seismic Evaluation Process Report in Response to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, December 18, 2014 (ADAMS Accession No. ML15005A074).

44. NRC letter to Duke Energy, "Brunswick Steam Electric Plant, Units 1 and 2 - Staff Review of Interim Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 (TAC Nos. MF5228 and MF5229)," November 19, 2015 (ADAMS Accession No. ML15313A245).

45. NRC letter to Duke Energy, "Brunswick Steam Electric Plant, Units 1 and 2 - Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard in Response to March 12, 2012, 50.54(f) Request for Information," April 20, 2017 (ADAMS Accession No. ML17107A277).

46. Exelon Generation Company 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).

47. NRC letter to Exelon Generation Company, 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).

48. NRC letter to Exelon Generation Company, 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).

49. Southern Nuclear Operating Company 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).

U.S. Nuclear Regulatory Commission Page 26 RA-23-0122

50. NRC letter to Southern Nuclear Operating Company, 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).

51. 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).

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

53. 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).

54. 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, April 30, 2020 (ADAMS Accession No. ML20076A194).

55. Exelon Generation Company (Peach Bottom) 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).

56. Southern Nuclear Operating Company letter to NRC, "Vogtle Electric Generating Plant -

Units 1 & 2 License Amendment Request to Incorporate Seismic Probabilistic Risk Assessment into the 10 CFR 50.69 Categorization Process Response to Request for Additional Information (RAIs 4-11), February 21, 2018 (ADAMS Accession No. ML18052B342).

57. 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 (WBN-TS-17-24),

November 29, 2018 (ADAMS Accession No. ML18334A363).

58. Exelon Generation Company (Calvert Cliffs) letter to NRC, Revised Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, letter dated July 19, 2019, August 5, 2019 (ADAMS Accession No. ML19217A143).

U.S. Nuclear Regulatory Commission Page 27 RA-23-0122

59. Exelon Generation Company (FitzPatrick) letter to NRC, Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, July 30, 2021 (ADAMS Accession No. ML21211A078).

60. Constellation Energy Generation letter to NRC, Supplemental Information No. 1 for James A. FitzPatrick Nuclear Power Plant to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. and 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power reactors," March 4, 2022 (ADAMS Accession No. ML22063A135).

61. NRC letter to Constellation Energy Generation, James A. Fitzpatrick Nuclear Power Plant - Issuance of Amendment No. 352 Re: Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, August 23, 2022 (ADAMS Accession No. ML22196A061).

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

63. 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, dated November 24, 2020 (ADAMS Accession No. ML20329A433).

U.S. Nuclear Regulatory Commission RA-23-0122 Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324 / Renewed License Nos. DPR-71 and DPR-62 License Amendment Request to Revise the 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, Categorization Process to Reflect an Alternative Seismic Approach Attachment Facility Operating License Markup (Unit Nos. 1 and 2)

Brunswick Unit 1 App. B-5 Amendment No. 305 Amendment Number 305 Additional Conditions Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 License Amendment No. 305 dated April 30, 2021.

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to the NRC dated November 24, 2020, prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above. The issues identified in the attachment will be addressed and any associated changes will be made prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

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).

Implementation Date Upon implementation of Amendment No. 305.

Replace with "UNIT 1 FOL INSERT" Upon implementation of Amendment No. [XXX]

Brunswick Unit 2 App. B-5 Amendment No. 333 Amendment Number 333 Additional Conditions Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 2 License Amendment No. 333 dated April 30, 2021.

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to the NRC dated November 24, 2020, prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

The issues identified in the attachment will be addressed and any associated changes will be made prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

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).

Implementation Date Upon implementation of Amendment No. 333.

Replace with "UNIT 2 FOL INSERT" Upon implementation of Amendment No. [XXX]

UNIT 1 FOL INSERT Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 License Amendment No. 305 dated April 30, 2021.

In addition, Duke Energy is approved to implement 10 CFR 50.69 using the alternative seismic approach for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs as described in Duke Energy letter RA-23-0122, dated August 17, 2023, as specified in Unit 1 License Amendment No. [XXX] dated [DATE].

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

UNIT 2 FOL INSERT Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 2 License Amendment No. 333 dated April 30, 2021.

In addition, Duke Energy is approved to implement 10 CFR 50.69 using the alternative seismic approach for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs as described in Duke Energy letter RA-23-0122, dated August 17, 2023, as specified in Unit 2 License Amendment No. [XXX] dated [DATE].

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