ML25213A180

Issuance of Amendment Nos. 287 and 280 Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors

ML25213A180
Person / Time
Site:	Dresden   (DPR-019, DPR-025)
Issue date:	08/07/2025
From:	Shilpa Arora Plant Licensing Branch III
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
Arora, S
References
EPID L-2024-LLA-0069
Download: ML25213A180 (33)

Text

August 7, 2025 Mr. David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - ISSUANCE OF AMENDMENT NOS. 287 AND 280 RE: ADOPTION OF 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2024-LLA-0069)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 287 to Renewed Facility Operating License No. DPR-19 and Amendment No. 280 to Renewed Facility Operating License No. DPR-25 for Dresden Nuclear Power Station, Units 2 and 3, respectively.

These amendments consist of changes to Renewed Facility Operating Licenses in response to your application dated May 28, 2024, as supplemented by your letters dated March 21, 2025, and April 23, 2025.

The amendments adopt Title 10 of the Code of Federal Regulations, Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's monthly Federal Register notice.

Sincerely,

/RA/

Surinder S. Arora, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-237 and 50-249

Enclosures:

1. Amendment No. 287 to DPR-19

2. Amendment No. 280 to DPR-25

3. Safety Evaluation cc: Listserv

CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 287 Renewed License No. DPR-19

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Constellation Energy Generation, LLC (the licensee), dated May 28, 2024, as supplemented by letters dated March 21, 2025, and April 23, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by the addition of paragraph 2.L to Renewed Facility Operating License No. DPR-19 which reads as follows:

L.

Constellation Energy Generation, LLC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire, seismic events; 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 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except external flooding and extreme winds and tornadoes; an external flood safe shutdown equipment list for external floods; and a high wind safe shutdown equipment list for extreme winds and tornadoes.

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 probabilistic risk assessment approach to seismic margins approach).

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 7, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.08.07 15:03:22 -04'00'

CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 280 Renewed License No. DPR-25

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Constellation Energy Generation, LLC (the licensee), dated May 28, 2024, as supplemented by letters dated March 21, 2025, and April 23, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by the addition of paragraph 3.FF to Renewed Facility Operating License No. DPR-25 which reads as follows:

FF Constellation Energy Generation, LLC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire, seismic events; 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 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except external flooding and extreme winds and tornadoes; an external flood safe shutdown equipment list for external floods; and a high wind safe shutdown equipment list for extreme winds and tornadoes.

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 probabilistic risk assessment approach to seismic margins approach).

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 7, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.08.07 15:03:51 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 287 AND 280 DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 Replace the following pages of the Renewed Facility Operating License with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

REMOVE INSERT License DPR-19 License DPR-19 Page 10A Page 10A License DPR-25 License DPR-25 Page 10B

-10A-Renewed License No. DPR-19 Amendment No. 287 K.

Adoption of Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extension Completion Times -RITSTF Initiative 4b Constellation is approved to implement TSTF-505, Revision 2, modifying the Technical Specifications requirements related to Completion Times (CTs) for Required Actions to provide the option to calculate a longer, risk-informed CT (RICT). The methodology for using the new Risk informed Completion Time Program is described in NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, which was approved by the NRC on May 17, 2007.

Constellation will complete the implementation items listed in of Constellation Energy Generation, LLC letter to the NRC dated May 8, 2024, prior to implementation of the RICT Program for these systems. All issues identified in Attachment 5 will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 3), and any findings will be resolved and reflected in the PRA of record prior to the implementation of the RICT Program.

L.

Constellation Energy Generation, LLC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire, seismic events; 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 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except external flooding and extreme winds and tornadoes; an external flood safe shutdown equipment list for external floods; and a high wind safe shutdown equipment list for extreme winds and tornadoes.

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 probabilistic risk assessment approach to seismic margins approach).

- 10B -

Renewed License No. DPR-25 Amendment No. 280 FF. Constellation Energy Generation, LLC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire, seismic events; 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 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except external flooding and extreme winds and tornadoes; an external flood safe shutdown equipment list for external floods; and a high wind safe shutdown equipment list for extreme winds and tornadoes.

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 probabilistic risk assessment approach to seismic margins approach).

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 287 AND 280 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 CONSTELLATION ENERGY GENERATION, LLC DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-237 AND 50-249

1.0 INTRODUCTION

By letter dated May 28, 2024 (Reference [1]), as supplemented by letters dated March 21, 2025 (Reference [2]), and April 23, 2025 (Reference [3]), Constellation Energy Generation, LLC (CEG, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for Dresden Nuclear Power Station, Units 2 and 3 (DNPS, Dresden). The amendments would allow the licensee to implement 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 licensee proposed the following license condition for DNPS Renewed Facility Operating License (RFOL) Nos. DPR-19 and DPR-25:

CEG is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire, seismic events; 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 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE [Individual Plant Examination of External Events] Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS [American Society of Mechanical Engineers (ASME) / American Nuclear Society] PRA Standard RA-Sa-2009 for other external hazards except external flooding and extreme winds and tornadoes; an external flood safe shutdown equipment list for external floods; and a high wind safe shutdown equipment list for extreme winds and tornadoes.

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 probabilistic risk assessment approach to seismic margins approach).

The NRC staff participated in a regulatory audit (Reference [4]) to ascertain the information needed to support its review of the application and to develop requests for additional information (RAIs), as needed. Following the regulatory audit, the licensee submitted a supplement letter dated March 21, 2025, that included additional information resulting from the audit. The NRC staff transmitted an RAI to the licensee by letter dated March 25, 2025 (Reference [5]). On April 23, 2025, the licensee submitted a response to the RAI. The NRC staff issued a regulatory audit report summary on May 20, 2025 (Reference [6]).

The supplemental letters dated March 21 and April 23, 2025, provided additional information that clarified the application, did not expand the scope of the application for 10 CFR 50.69 as originally submitted, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 6, 2024 (89 FR 63991).

2.0 REGULATORY EVALUATION

2.1 Applicable Regulations The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on an integrated and systematic risk-informed process that includes several approaches and methods for categorizing SSCs according to their safety significance1. Special treatment refers to those requirements that provide increased assurance (beyond normal industry practices) that SSCs will perform their design-basis functions. For SSCs with a function that is categorized as low-safety-significant (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs with a function determined to be high-safety-significant (HSS), requirements may not be changed.

The regulation, 10 CFR 50.69, contains requirements regarding how a licensee categorizes SSCs using a risk-informed process; adjusts treatment requirements consistent with the relative significance of the SSC; and manages the process over the lifetime of the plant. A risk-informed process is employed to determine the safety significance of SSCs and assign each into one of four RISC categories.

SSC categorization does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility.

Instead, 10 CFR 50.69 enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained and may be enhanced. Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative, risk-informed approach to treatment that provides a reasonable level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on HSS equipment.

1 Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, May 2006, describes the SSC categorization process in its entirety as an overarching approach that includes multiple approaches and methods identified for a PRA hazard and non-PRA methods.

2.2 Regulatory Guidance The NRC staff considered the following regulatory guidance during its review of the proposed changes:

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference [7])

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference [8])

RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference [9])

RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (Reference [10])

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking (Reference [11])

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP) Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance (Reference [12])

The licensees submittal cites RG 1.174, Revision 2, as a guidance document applicable to the proposed change. RG 1.174 has been updated to Revision 3. The update does not include any technical changes that impact the consistency with RG 1.201 and implementation of the SSC categorization program; therefore, the NRC staff finds the updated revision to RG 1.174 also applicable for use in the licensees adoption of 10 CFR 50.69 and will be the revision referred to in the remainder of the safety evaluation (SE).

NRC-Endorsed Guidance:

The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference [13]), as endorsed by RG 1.201 for trial use with clarifications and describes a process that the NRC staff considers acceptable for complying with 10 CFR 50.69. This process determines the safety significance of SSCs and categorizes them into one of four RISC categories defined in 10 CFR 50.69.

Sections 2 through 10 of NEI 00-04 describe the following steps/elements of the SSC categorization process for meeting the requirements of 10 CFR 50.69:

Sections 3.2 and 5.1 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i).

Sections 3, 4, 5, and 7 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii).

Section 6 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iii).

Section 8 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv).

Section 2 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v).

Sections 9 and 10 provide specific guidance corresponding to 10 CFR 50.69(c)(2).

Additionally, Section 11 of NEI 00-04 provides guidance on program documentation and change control related to the requirements of 10 CFR 50.69(f). Section 12 of NEI 00-04 provides guidance on the periodic review related to the requirements in 10 CFR 50.69(e). Maintaining change control and periodic review provides confidence that all aspects of the program reasonably reflect the current as-built, as-operated plant configuration and applicable plant and industry operational experience as required by 10 CFR 50.69 (c)(1)(ii).

3.0 TECHNICAL EVALUATION

3.1 Method of NRC Staff Review An acceptable approach for making risk-informed decisions about proposed changes, including both permanent and temporary changes, is to show that the proposed changes to the licensing basis meet the five key principles of risk-informed decision-making stated in Section C of RG 1.174, Revision 3:

Principle 1:

The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption.

Principle 2:

The proposed licensing basis change is consistent with the defense-in-depth philosophy.

Principle 3:

The proposed licensing basis change maintains sufficient safety margins.

Principle 4:

When proposed licensing basis changes result in an increase in risk, the increases should be small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants.

Principle 5:

The impact of the proposed licensing basis change should be monitored using performance measurement strategies.

The engineering evaluations below address the five key principles of RG 1.174, Revision 3.

3.2 Overview of Categorization Process Paragraph 50.69(c) of 10 CFR requires licensees to use an integrated decision-making process to categorize safety-related and non-safety-related SSCs according to the safety significance of the functions they perform. They are placed into one of the following four RISC categories:

RISC-1: Safety-related SSCs that perform safety significant functions2 RISC-2: Non-safety-related SSCs that perform safety significant functions RISC-3: Safety-related SSCs that perform low safety significant functions RISC-4: Non-safety-related SSCs that perform low safety significant functions 2 NEI 00-04, Revision 0, uses the term high-safety-significant to refer to SSCs that perform safety-significant functions. The NRC understands HSS to have the same meaning as safety-significant as used in 10 CFR 50.69, which applies to RISC-1 and RISC-2 SSCs.

The SSCs have functions that are HSS or LSS, and they are classified accordingly. For SSCs that perform HSS functions (i.e., RISC-1 and RISC-2 SSCs), 10 CFR 50.69 maintains current regulatory requirements for special treatment, that is, all existing special treatment requirements continue to apply. In addition, 10 CFR 50.69(d)(1) requires licensees to ensure that RISC-1 and RISC-2 SSCs perform their functions consistent with the categorization process assumptions by evaluating treatment being applied to these SSCs to ensure that it supports the key assumptions in the categorization process that relate to their assumed performance. For SSCs that perform LSS functions, licensees may implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2). For RISC-3 SSCs, licensees may replace special treatment requirements with an alternative treatment approach that meets 10 CFR 50.69(d)(2). For RISC4 SSCs, 10 CFR 50.69 does not impose new treatment requirements.

Paragraph 50.69(b)(3) of 10 CFR states that the Commission will approve a licensees implementation of this section by issuance of a license amendment if the Commission determines that the categorization process satisfies the requirements of 10 CFR 50.69(c). As stated in 10 CFR 50.69(b), after the NRC approves an application for a license amendment, a licensee may voluntarily comply with 10 CFR 50.69 as an alternative to compliance with the following requirements for LSS SSCs:

(i) 10 CFR Part 21 (ii) the specified portion of 10 CFR 50.46a(b)

(iii) 10 CFR 50.49 (iv) 10 CFR 50.55(e)

(v) specified requirements of 10 CFR 50.55a (vi) 10 CFR 50.65, except for paragraph (a)(4)

(vii) 10 CFR 50.72 (viii) 10 CFR 50.73 (ix)

Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50 (x) specified requirements for containment leakage testing (xi) specified requirements of Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, to 10 CFR Part 100 The NRC staff reviewed the licensees SSC categorization process against the categorization process described in NEI 00-04, and the acceptability of the licensees PRA for use in the application of the 10 CFR 50.69 categorization process. The NRC staffs review, as documented in this SE, used the framework provided in RG 1.174, Revision 3, and RG 1.201, Revision 1.

Section 2 of NEI 00-04 states that the categorization process includes the following eight primary steps:

1.

Assembly of Plant-Specific Inputs

2.

System Engineering Assessment

3.

Component Safety Significance Assessment

4.

Defense-in-Depth Assessment

5.

Preliminary Engineering Categorization of Functions

6.

Risk Sensitivity Study

7.

Integrated Decision-Making Panel Review and Approval

8.

SSC Categorization 3.3 Traditional Engineering Evaluation The traditional engineering evaluation below addresses the first three key principles identified in RG 1.174, Revision 3 and are pertinent to: (1) compliance with current regulations, (2) evaluation of defense-in-depth, and (3) evaluation of safety margins.

3.3.1 Key Principle 1: Licensing Bases Change Meets the Current Regulations In Section 3.1.1, Overall Categorization Process, of the LAR dated May 28, 2024, the licensee states that it will implement the risk categorization process in accordance with NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1. The licensee provided further discussion of specific elements within the 10 CFR 50.69 categorization process that are delineated in the endorsed guidance of NEI 00-04.

The regulatory requirements in 10 CFR 50.69 and 10 CFR Part 50, Appendix B, as well as the monitoring outlined in NEI 00-04, will ensure that the SSC functions continue to be met, that any performance deficiencies will be identified, and that appropriate corrective actions will be taken.

The NRC staff finds that the licensees SSC categorization program includes the appropriate steps/elements prescribed in NEI 00-04 to assure that SSCs are appropriately categorized, consistent with 10 CFR 50.69. Therefore, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the first key principle for risk-informed decision-making identified in RG 1.174, Revision 3.

3.3.2 Key Principle 2: Licensing Basis Change is Consistent with the Defense-In-Depth Philosophy In RG 1.174, Revision 3, the NRC identified the following considerations used for evaluating how the licensing basis change is maintained for the defense-in-depth philosophy:

Preserve a reasonable balance among the layers of defense.

Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

Preserve adequate defense against potential common-cause failures.

Maintain multiple fission product barriers.

Preserve sufficient defense against human errors.

Continue to meet the intent of the plants design criteria.

In Section 3.1.1 of the LAR enclosure dated May 28, 2024, the licensee clarified that, consistent with the guidance in NEI 00-04, it would require an SSC categorized as HSS based on the defense-in-depth assessment to be categorized HSS per the final categorization, and that cannot be changed by the integrated decision-making panel (IDP).

The NRC staff finds that the licensees process is consistent with the NRC-endorsed guidance in NEI 00-04 and concludes that the proposed change is consistent with the defense-in-depth philosophy. For this reason, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the second key principle for risk-informed decision-making identified in RG 1.174, Revision 3, and fulfills the requirement of 10 CFR 50.69(c)(1)(iii) that defense-in-depth be maintained.

3.3.3 Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins The third key risk-informed principle in RG 1.174 states that the licensing basis change should maintain sufficient safety margins. The engineering evaluation that will be conducted by the licensee under 10 CFR 50.69 for SSC categorization will assess the design function(s) and risk significance of SSCs to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring the categorization of the SSC does not adversely affect any assumptions or inputs to the safety analysis; or, if such inputs are affected, providing justification that sufficient safety margin will continue to exist.

The NRC staff notes that the design-basis functions of SSCs, as described in the plants licensing basis, including the DNPS Updated Final Safety Analysis Report (UFSAR) and plant technical specification (TS) bases, do not change and the safety margins described should continue to be met. Similarly, there is no impact to the safety analysis acceptance criteria, as described in the plant licensing basis. On this basis, the NRC staff concludes that safety margins are maintained by the proposed methodology, and the third key safety principle identified in RG 1.174, Revision 3 is satisfied.

In Section 2.2, Reason for Proposed Change, of the LAR enclosure dated May 28, 2024, the licensee states, The safety functions [in the categorization process] include the design basis functions, as well as functions credited for severe accidents (including external events).

Section 3.1.1 of the LAR enclosure summarizes the different hazards and plant states for which functional and risk significant information will be collected. In the same section, the licensee confirmed that the SSC categorization process documentation will include, among other items, system functions, identified and categorized with the associated bases, and mapping of components to supported function(s).

The NRC staff finds that the process described in the LAR is consistent with NEI 00-04, as endorsed by the NRC in RG 1.201, Revision 1. Therefore, it meets the requirements set forth in 10 CFR 50.69(c)(1)(ii) and 10 CFR 50.69(c)(1)(iv).

3.3.4 Key Principle 4: Change in Risk is Consistent with Safety Goals The risk-informed considerations prescribed in NEI 00-04 address the fourth key principle of risk-informed decision-making identified in RG 1.174, Revision 3. A summary of how the licensees SSC categorization process is consistent with the guidance and methodology in NEI 00-04, Revision 0, and RG 1.201, Revision 1, is provided in the sections below. The NRC staff acknowledges that elements of the categorization process are not always performed in chronological order and may be performed in parallel.

3.3.4.1 Probabilistic Risk Assessment Scope of the PRA The Dresden PRA is comprised of full-power, internal events PRA (IEPRA), including internal floods, fire PRA (FPRA), and seismic PRA (SPRA), which evaluates the core damage frequency (CDF) and large early release frequency (LERF) risk metrics. The licensee also states that the PRA models have been independently peer reviewed and there are no PRA upgrades that have not been peer reviewed.

In Section 3.3, PRA Review Process Results (10 CFR 50.69(b)(2)(iii)), the licensee provided a description of the peer reviews and the associated peer review Facts and Observations (F&O) findings closure reviews performed for the PRA models. F&O closure reviews were performed using the process documented in the NEI letter to the NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-13, Close-out of Facts and Observations (F&Os) (Appendix X) dated February 21, 2017 (Reference [14]), as accepted by the NRC in its letter dated May 3, 2017 (Reference [15]).

Aspects considered by the staff to evaluate the scope of the PRA include: (1) peer-review history and results, (2) the Appendix X, Independent Assessment process, (3) credit for the diverse and flexible coping strategy (FLEX) in the PRA, and (4) assessment of assumptions and approximations.

The information provided in the LAR dated May 28, 2024, is sufficient to support the NRC staffs review of the DNPS PRAs, and; therefore, the NRC staff finds that it meets the requirements of 10 CFR 50.69(b)(2)(iii).

Internal Events PRA (includes internal floods) and Internal Fires PRA By letter dated May 28, 2024, in the LAR application to adopt 10 CFR 50.69 for DNPS, Units 2 and 3, the licensee states that the PRA models described within this LAR are the same as those described within the CEG submittal dated May 8, 2024, to support DNPS adoption to Request to Revise Technical Specifications to Adopt Risk Informed Completion Times [Technical Specifications Task Force (TSTF)]-505, Revision 2, Provide Risk-Informed Extended Completion Times - [Risk-Informed Technical Specifications Task Force (RITSTF)] Initiative 4b, and TSTF-591, Revise Risk Informed Completion Time (RICT) Program, (Reference [16]).

The risk-informed applications for 10 CFR 50.69, TSTF-505, and TSTF-591 use the same technical supporting requirements from the endorsed ASME/ANS PRA Standard (Reference

[17]) to determine PRA acceptability. Previous staff review for DNPS issuance of the SE for the adoption of TSTF-505 and TSTF-591 (Reference [18]) concluded the IEPRA, including internal floods, and FPRA, are acceptable and consistent with RG 1.200, Revision 2.

Therefore, the NRC staff finds that for the 10 CFR 50.69 risk-informed LAR, the DNPS IEPRA (including internal floods) and FPRA were appropriately peer reviewed, including closure of F&Os consistent with RG 1.200, Revision 2, and the PRA assumptions and sources of uncertainty were appropriately identified consistent with NUREG-1855.

External Events: Seismic PRA In accordance with sections 7-1.2 and 8-1.2 of the PRA standard, it is assumed that full-scope, internal events, at-power, Level 1 and Level 2 LERF PRAs exist and that those PRAs are used as the basis for the SPRA. Therefore, the technical acceptability of the IEPRA model used as the foundation for the SPRA is an important consideration.

Section 3.3 of the enclosure to the LAR states that a focused-scope peer review was performed in June 2023 on the IEPRA model, which was after the SPRA had a F&O closure review in June 2022. In the LAR supplement response dated March 21, 2025, to APLC question 1.b, the licensee included an evaluation of IEPRA changes after June 2022 on the SPRA F&O closure review and results. Based on its review of the IEPRA model and evaluation of IEPRA change impacts on the SPRA, the NRC staff finds that the IEPRA is technically acceptable to be used as the foundation for the SPRA.

In Section 3.2.3 of the enclosure to the LAR dated May 28, 2024, the licensee stated that the proposed categorization process will use a peer-reviewed, plant-specific SPRA model. The NRC staffs review of the technical acceptability of the SPRA model for this application is discussed below.

Seismic PRA Peer-Review History The NRC staffs review of the licensees SPRA was based on the results of the peer review and the associated F&O closure review for closure of F&Os described in LAR Section 3.3 and attachment 3. The last full-scope peer review of the SPRA was performed in January 2019 against the SPRA requirements in ASME/ANS RA-Sb-2013 (Reference [19]), also known as Addendum B of the 2008 ASME/ANS PRA standard. RG 1.200, Revision 2, endorses ASME/ANS PRA Standard RA-Sa-2009, also known as Addendum A of the 2008 ASME/ANS PRA standard, but does not endorse Addendum B.

In the LAR supplement, response to APLC question 2, the licensee demonstrated that the supporting requirements (SRs) in part 5 of Addendum B are consistent with those in Addendum A. Based on its review of the licensees comparison of SRs of part 5 of Addendum B of the PRA standard compared to those in Addendum A, the NRC staff finds the licensees use of Addendum B to be an acceptable alternative to the NRC-endorsed approach for this application because it adequately addresses the technical elements for the development of a SPRA.

The SPRA peer review and F&O closure review used NEI 12-13 (Reference [20]) but did not explicitly address the NRC letter dated March 7, 2018, regarding external hazard peer reviews (Reference [21]). In the LAR supplement, response to APLC question 1.a, the licensee stated that the Dresden SPRA 2019 peer review and 2022 F&O closure processes are compliant with the NRC clarifications on NEI 12-13. Based on the information provided by the licensee as well as the documented comparison of the NEI 12-13 comments with the SPRA peer review and F&O closure processes, the NRC staff finds the performance of the SPRA peer review and F&O closure process to be acceptable.

The NRC staff reviewed the SPRA peer review results and the licensee's resolution of the results and finds that the Dresden SPRA was appropriately peer-reviewed consistent with RG 1.200, Revision 2, and the F&Os have been closed using an NRC-approved approach.

Treatment of the Key Assumptions and Sources of Uncertainty NUREG-1855, Revision 1, provides guidance regarding how to address PRA uncertainties to ensure the risk-informed decision is in the context of the application for the decision under consideration. The licensee confirmed that sensitivity studies will be performed consistent with the NEI 00-04 guidance. In accordance with NEI 00-04, the results of the sensitivity studies are given to the IDP for consideration in the final risk characterization for components initially classified as LSS that may be reclassified to HSS.

In Table A6 of the LAR, and Attachment 2, Section 2.1 of the supplement dated March 21, 2025, the licensee identified two key sources of uncertainty relative to the FLEX equipment: the FLEX and Hardened Containment Vent System (HCVS) and the BlackStarTech portable carts equipment. The licensee acknowledged there did not exist industry-approved data for the FLEX equipment, in addition to human failures events associated with the alignment of the equipment.

For the FLEX equipment key sources of uncertainty, the licensee performed sensitivities using the IEPRA model removing the credit and, in the supplement, provided additional studies using the IEPRA and FPRA to demonstrate the sensitivity results were valid. The updated results continued to demonstrate no significant impact on SSC categorization.

In Table A6 of the LAR, specific to the key assumption associated with Instrument Air (IA) system recovery, the licensee stated that the model uses data from NSAC-161 to provide a basis to support the restoration of IA or its support systems. In the LAR supplement, the licensee stated that NSAC-161 was used in the original development of the PRA and should not have been included in the LAR. The licensee updated the current PRA model to use a more conservative estimate. The licensee confirmed that the conclusion in LAR Table A6 that the CDF and LERF are relatively insensitive to the failure probabilities associated with the recovery of IA, specific to support of venting remains valid.

Based on this review, the licensee determined that no additional sensitivity analyses are required to address DNPS PRA model assumptions or sources of uncertainty, as discussed in Section 3.2.7 of the LAR, as supplemented. The NRC staff finds that the licensee performed sensitivity studies consistent with NEI 00-04 guidance to address the identified key assumptions and sources of uncertainty specific to the SSC categorization program.

In addition, the NRC staff recognizes that the licensee will perform routine PRA changes and updates to ensure the PRA continually reflects the as-built, as-operated plant, in addition to changes made to the PRA to support the context of the analysis being performed (i.e.,

sensitivities). Paragraph 50.69(e) and (f) stipulates the process for feedback and adjustment to ensure configuration control is maintained for these routine changes and updates to the PRA(s).

PRA Importance Measures and Integrated Importance Measures Pursuant to 10 CFR 50.69(c)(1)(ii), the licensees SSC characterization process must determine SSC functional importance using an integrated, systematic process for addressing internal and external initiating events. NEI 00-04, Section 5, provides guidance on the risk importance assessment process. The scope of modeled hazards for DNPS includes the IEPRA (includes internal floods), FPRA. and SPRA. The NRC staff reviewed the LAR and finds that the licensees use and treatment of importance measures is consistent with the guidance in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1.

PRA Acceptability Conclusions Pursuant to 10 CFR 50.69(c)(1)(i), the categorization process must consider results and insights from a plant-specific PRA. The use of the IEPRA, FPRA, and SPRA to support SSC categorization is endorsed by RG 1.201, Revision 1. The PRAs must be acceptable to support the categorization process and must be subjected to a peer-review process assessed against a standard that is endorsed by the NRC. Revision 2 of RG 1.200 provides guidance for determining the acceptability of the PRA by comparing the PRA to the relevant parts of the ASME/ANS 2009 Standard using a peer-review process.

The licensee has subjected the IEPRA, FPRA, and SPRA to the peer-review processes and submitted the results of the peer review. The NRC staff reviewed the peer-review history (which included the results and findings), the licensee's resolution of peer-review findings, and the identification and disposition of key assumptions and sources of uncertainty. The NRC staff concludes that (1) the licensee's IEPRA, FPRA, and SPRA are acceptable to support the categorization of SSCs using the process endorsed by the NRC in RG 1.201, Revision 1, and (2) the key assumptions for the PRAs have been identified consistent with the guidance in RG 1.200, Revision 2 and NUREG-1855 (References [10], [8], and [11], respectively), as applicable, and addressed appropriately for this application.

The NRC staff finds that the licensee provided the required information, and the IEPRA (includes internal floods), FPRA, and SPRA are acceptable and therefore meet the requirements set forth in paragraphs 50.69(c)(1)(i) and (ii) of 10 CFR 50.69.

Evaluation of the Use of Non-PRA Methods in SSC Categorization The licensees categorization process uses the following non-PRA methods, respectively:

Screening analysis performed for the IPEEE for external hazards (Reference [22]),

updated to reflect part 6 of ASME PRA Standard RA-Sa-2009 Safe shutdown equipment lists for equipment needed to achieve and maintain safe shutdown of the reactor for unscreened external hazards (e.g., high winds and external flooding)

Safe shutdown risk management program consistent with NUMARC 91-06 (Reference [23])

ANO-2 passive categorization for passive components (Reference [24])

The NRC staff's review of these methods is discussed below.

External Flooding Approach The licensee determined that the combined effects riverine flood mechanism does not screen out for the Dresden categorization process and the local intense precipitation (LIP) flood scenario can be screened by crediting various doors for flood protection. In Section 3.2.4.1 of the LAR, the licensee stated that the process used for the external flood hazard safety significance involves an external flood safe shutdown equipment list (XFSSEL) that is comprised of the key SSCs needed to achieve and maintain safe shutdown of the reactor for external riverine floods producing water surface elevations higher than plant grade of 517.5 feet mean sea level (MSL). The licensee further stated that these key SSCs would be classified as HSS during the 10 CFR 50.69 categorization process regardless of their flood damage susceptibility or frequency of challenge. Additionally, the licensee stated that there would be no reliance on operator actions in the determination if an SSC should be assigned to the XFSSEL.

In Section 3.2.4.1 of the LAR, the licensee provided details regarding the methodology to be used to develop the XFSSEL, which includes the NEI 00-04 component to function mapping process for safe shutdown functions and six criteria used to screen out SSCs. The licensee made one change to the third of six criteria in its RAI response dated April 23, 2025 to the Dresden TSTF-505 application. This change is specific to the Dresden 50.69 applicationand is highlighted as such in the RAI responsebut the change was made during an adjustment to how external flooding risk is evaluated as part of the concurrent TSTF-505 application. The six criteria included for screening SSCs not otherwise credited for mitigating the effects of an external riverine flood higher than 517.5 feet MSL in elevation include SSCs not powered by emergency on-site alternating current (AC) sources, SSCs not required to function during or after a loss of offsite AC power event, SSCs in systems that are assumed unavailable during or following an external flooding event, SSCs outside a Category I structure not protected against external floods, SSCs that only perform a passive safety function during an external flooding event, and SSCs determined by Dresden operations to not be part of the external flooding mitigation strategy. The remaining SSCs not screened out will be identified as HSS SSCs on the XFSSEL.

In the LAR supplement dated March 21, 2025, Section 4.3, Question 03 - Equipment Lists, the licensee clarified that all six criteria must be met in order to categorize an SSC as LSS with respect to the external flooding hazard. The methodology is conservative because all six criteria must be met for an SSC on the XFSSEL to be screened as LSS and any SSC screened as LSS must be well justified.

In the LAR supplement dated March 21, 2025, Section 4.4, Question 04 - External Flooding Screening, the licensee corrected statements made in the original LAR. In the original LAR dated May 28, 2024, a reactor trip was provided as an example for an SSC not required to function during or after a loss of offsite power event. The licensee stated that the example of a reactor trip being screened out should be removed as it does not apply to the criteria.

Additionally, the licensee corrected a text error, stating that there are 18 doors necessary for screening out the LIP flood scenario. The staff agrees with the corrections, which are necessary to correctly implement the categorization of SSCs related to the external flooding hazards.

Therefore, the NRC staff concludes that the approach for developing the XFSSEL that will be used for categorizing SSCs for external flooding is conservative because (1) the approach ensures that SSCs that are credited to achieve and maintain the capability for safe shutdown are retained as HSS; (2) SSCs identified as HSS by non-PRA methods for external events may not be re-categorized by the IDP, as stated in NEI 00-04 and the LAR; (3) the external flooding approach does not rely on operator actions in determination of SSC safety categorization; and (4) the SSCs that fulfill the external flood safe shutdown functions, as well as any external flood barriers that are credited with protecting equipment that fulfills a XFSSEL function, will be identified as candidate HSS for the system being categorized regardless of flood damage susceptibility or frequency of challenge. In Table APLC 3-1 of the LAR supplement dated March 21, 2025, the licensee provided an implementation item to develop the DNPS XFSSEL prior to SSC categorization. The NRC staff finds the use of the DNPS XFSSEL cited in the proposed license condition of Section 4.0 of this SE acceptable for SSC categorization.

High Winds and Tornado/Tornado Missiles Approach The licensee determined that the straight winds and tornado/tornado missiles hazards do not screen out for the Dresdens categorization process. In Section 3.2.4.2 of the LAR, the licensee stated that Dresden had previously developed a tornado safe shutdown equipment list (TSSEL) in response to Regulatory Issue Summary (RIS) 2015-06, Tornado Missile Protection, (Reference [25]) which will be used as the high winds safe shutdown equipment list (HWSSEL) for SSC categorization under 10 CFR 50.69. The basis for use of the TSSEL is the straight and tornado wind components of the high wind and tornado/tornado missile risk are driven mainly by wind pressure and random failures of the three diesel generators since a loss of offsite power is assumed. The licensee further stated that these key SSCs would be classified as HSS during the 10 CFR 50.69 categorization process regardless of their high wind and tornado/tornado missile damage susceptibility or frequency of challenge. Additionally, the licensee stated that there is no reliance on operator actions in the determination if an SSC should be assigned to the HWSSEL.

In section 3.2.4.2 of the LAR, the licensee stated that the TSSEL addressed the safe shutdown functions of reactivity control, inventory control, pressure control, and decay heat removal capability. In the LAR supplement dated March 21, 2025, section 4.3, the licensee stated that, although the power availability function is not addressed explicitly in the TSSEL, it is implied that emergency power availability (alternating current (AC) and direct current (DC)) is relied upon for the safe shutdown strategy. Since offsite power is assumed lost in the TSSEL, the emergency diesel generators are included in the list of equipment in the TSSEL. Additionally, the licensee stated that the NEI 00-04 component to function mapping process will be applied during categorization of systems, which will include the power availability safe shutdown function.

Therefore, the NRC staff finds that the HWSSEL approach to categorizing SSCs for the high winds and tornado/tornado missiles hazards is conservative because (1) the approach ensures that SSCs credited to achieve and maintain the capability for safe shutdown are retained as HSS; (2) SSCs identified as HSS by non-PRA methods for external events may not be re-categorized by the IDP, as stated in NEI 00-04 and the LAR; (3) the high winds approach does not rely on operator actions in determination of SSC safety categorization; and (4) the SSCs that fulfill the wind pressure/missile hazard safe shutdown functions, as well as any high winds or missile barriers that are credited with protecting equipment that fulfills a HWSSEL function, will be identified as candidate HSS for the system being categorized regardless of their tornado damage susceptibility or frequency of challenge.

Methods for Assessing Other External Hazards This hazard category includes all non-seismic external hazards except for fire, extreme winds and tornadoes, and external flooding for Units 2 and 3. In Section 3.2.4.3 of the LAR dated May 28, 2024, the licensee stated, in part, that all other external hazards (i.e., not seismic, fire, external flooding, or extreme winds and tornado hazards) were screened from applicability to Dresden per a plant-specific evaluation in accordance with Generic Letter (GL) 88-20, Individual Plant Examination for Severe Accident Vulnerabilities, Supplement 4, dated June 1991, and updated to use the criteria in ASME/ANS PRA Standard RA-Sa-2009.

In Attachment 4 of the LAR, the licensee provided the results of the plant-specific evaluation that assessed the IPEEE results using endorsed criteria in the ASME/ANS PRA Standard RA-Sa-2009 and current plant hazard information. The NRC staff notes this plant-specific evaluation and its results were not peer reviewed against part 6 of the ASME/ANS PRA Standard Ra-SA-2009, as endorsed in RG 1.200, Revision 2.

In the LAR supplement dated March 21, 2025, Section 4.5, External Hazards Screening, the licensee stated that a different criterion is also valid for screening the biological fouling external hazard; namely, that it is an event that develops slowly, allowing adequate time to eliminate the threat. The licensee based its conclusion on the long amount of time over which Asiatic clams foul the safety-related service water systems, and Dresdens infestation trending program implemented to monitor this external hazard.

In the LAR supplement dated March 21, 2025, Section 4.5, the licensee provided justification for the amount of time necessary to mitigate a toxic gas external hazard. The immediate action of the procedure for smoke, noxious fumes, or airborne contaminants in the control room is to don a self-contained breathing apparatus (SCBA), and this procedure is readily available in the control room. The licensee stated that the Dresdens UFSAR (Reference [26]) includes a statement that 2 minutes is considered sufficient time for donning an SCBA per RG 1.78, Revision 0, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release (Reference [27]). Additionally, the abnormal operating procedure used to place the control room HVAC in an isolated recirculation alignment was referenced, which involves the operation of a single switch in the main control room.

There were two errors in the original LAR that the licensee addressed in the LAR supplement dated March 21, 2025, Section 4.5. One error was associated with a language discrepancy between Dresdens 50.69 and TSTF-505 LARs with respect to the external hazard screening tables for the low lake or river water level. The other error was associated with the wrong screening criteria referenced for the high summer temperature external hazard. The licensee adjusted the external hazard screening table as appropriate in the LAR supplement to correct the errors.

In summary, the use of the Dresden IPEEE results described by the licensee in the LAR dated May 28, 2024, supplemental information provided in the LAR supplement dated March 21, 2025, and the licensee's assessment of the other external hazards (i.e., high winds, tornadoes, and external flooding) is consistent with section 5 of NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1 (References [13] and [10], respectively). The NRC staff concludes that the licensee's treatment of other external hazards is acceptable and meets 10 CFR 50.69(c)(1)(ii).

Shutdown Risk Consistent with the guidance in NEI 00-04, Revision 0, the licensee proposed using the shutdown safety assessment based on NUMARC 91-06 (References [13] and [23],

respectively). NUMARC 91-06 provides considerations for maintaining defense-in-depth for the five key safety functions during shutdown, namely, decay heat removal capability, inventory control, power availability, reactivity control, and containment-primary/secondary. NUMARC 91-06 also specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. This is accomplished by designating a running and an alternative system/train to accomplish the given key safety function.

The use of NUMARC 91-06 described by the licensee in its submittal is consistent with the guidance in NEI 00-04, Revision 0, as endorsed by the NRC in RG 1.201, Revision 1 (References [13] and [10], respectively). The approach uses an integrated and systematic process to identify HSS components, consistent with the shutdown evaluation process.

Therefore, the NRC staff finds that the licensee's use of NUMARC 91-06 is acceptable, and it meets the requirements set forth in 10 CFR 50.69(c)(1)(ii).

Component Safety Significance Assessment for Passive Components Passive components are not modeled in the PRA; therefore, a different assessment method is necessary to assess the safety significance of these components. Passive components are those components having only a pressure retaining function. This process also addresses the passive function of active components such as the pressure/liquid retention of the body of a motor-operated valve.

In Section 3.1.2 of the LAR dated May 28, 2024, the licensee proposed using a categorization method for passive components not cited in NEI 00-04, Revision 0, or RG 1.201, Revision 1 (References [13] and [10], respectively), but was approved by the NRC for ANO-2. The ANO-2 methodology is a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure retaining items and their associated supports (exclusive of Class CC and MC items), using a modification of the ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1 (Reference [29]). The ANO-2 methodology relies on the conditional core damage and large early release probabilities associated with pipe ruptures.

Safety significance is generally measured by the frequency and the consequence of, in this case, pipe ruptures. Treatment requirements (including repair/replacement) only affect the frequency of passive component failure. Categorizing solely based on consequences, which measures the safety significance of the pipe given that it ruptures, is conservative compared to including the rupture frequency in the categorization. The categorization will not be affected by changes in frequency arising from changes to the treatment. Therefore, the NRC staff finds that the use of the repair/replacement methodology is acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs.

In Section 3.1.2, of the enclosure to its LAR, the licensee stated, in part, the passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the ANO2-R&R-004 for the passive categorization of Class 2, 3, and non-class components. Consistent with ANO2-R&R-004, Class 1 pressure retaining SSCs, as well as supports, in the scope of the system being categorized will be assigned HSS and cannot be changed by the IDP. That is, the ANO-2 repair/replacement methodology does not allow a Class 1 pressure retaining SSC to be recategorized from HSS to LSS. Therefore, the NRC staff finds the licensee's proposed approach for passive categorization is acceptable for the 10 CFR 50.69 SSC categorization process.

Assembly of Plant-Specific Inputs The licensees risk categorization process uses PRAs to assess risks from internal events (including internal floods), internal fire, and seismic hazards. Non-PRA methods are used to assess shutdown safety, passive component risk, and the significance of external hazards.

The process used by the licensee is described in the LAR dated May 28, 2024, as supplemented by letter dated March 21, 2025. The process for collecting and organizing information at the system level for defining boundaries, functions, and components is consistent with NEI 00-04. Because the process is consistent with NEI 00-04 as clarified and endorsed in RG 1.201, Revision 1, the NRC staff finds that the process meets the requirements set forth in 10 CFR 50.69(c)(1)(v).

Risk Sensitivity Study (NEI 00-04, Section 8)

Section 3.1.1 of the LAR dated May 28, 2024, states that an unreliability factor of three will be used for the sensitivity studies described in Section 8, Risk Sensitivity Study, of NEI 00-04, Revision 0. Section 3.2.7 of the LAR further confirms that a cumulative sensitivity study will be performed where the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in PRAs for all systems that have been categorized are increased by a factor of three. The NRC staff finds the application of a factor of three for the sensitivities is consistent with the guidance in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1 (References [13] and [10], respectively).

In Section 3.1.1 of the LAR, for the Overall Categorization Process, CEG specifically noted that the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(iv). This sensitivity study together with the periodic review process, discussed in Section 3.5.2 of this SE, ensures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of the categorized components are detected and addressed before reaching the rate assumed in the sensitivity study. The NRC staff finds that the licensee will perform the risk sensitivity study consistent with the guidance in section 8 of NEI 00-04, Revision 0, and, therefore, will ensure that the potential cumulative risk increase from the categorization is maintained acceptably low, as required by 10 CFR 50.69(c)(1)(iv).

Integrated Decision-Making Appendix B of SRP Chapter 19, Section 19.2, provides guidance and the staffs expectations for the licensees integrated decision-making process. The appendix states in part, [r]isk-informed applications are expected to require a process to integrate traditional engineering and probabilistic considerations to form the basis for acceptance. NEI 00-04 guidance identifies two steps in the categorization process: (1) Preliminary Engineering Categorization of Function and (2) IDP Review and Approval that are responsible for the integrated assessment of the traditional engineering analyses and the risk results from the PRA and non-PRA assessments that are performed to make a determination and approval of the safety significance of the SSC for categorization. The staff review of the two steps to ensure the processes is well-defined, systematic, repeatable, and scrutable are provided as follows:

Preliminary Engineering Categorization of Function (NEI 00-04, Section 7)

In Section 3.1.1 of the LAR dated May 28, 2024, the licensee acknowledged the NRC staffs clarification of NEI 00-04 that if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS. The licensee also stated, Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The IDP must intervene to assign any of these HSS function components to LSS.

The NRC staff finds that the above description provided by the licensee for the preliminary categorization of functions is consistent with NEI 00-04 and is therefore acceptable.

IDP Review and Approval (NEI 00-04, Sections 9 and 10)

In Section 3.1.1 of the LAR dated May 28, 2024, the licensee states:

The IDP will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) engineering, system engineering, safety analysis, and probabilistic risk assessment. At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA.

The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address, at a minimum, (1) the purpose of the categorization; (2) present treatment requirements for SSCs including requirements for design basis events; (3) PRA fundamentals; (4) details of the plant-specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and (5) the defense-in-depth philosophy and requirements to maintain this philosophy.

In Table APLC 3-1 of the LAR supplement dated March 21, 2025, the licensee identified an implementation item that the IDP makeup will include a member with external flood expertise for any system categorized at Dresden. Furthermore, the guidance in NEI 00-04, as endorsed by the NRC in RG 1.201, provides confidence that the IDP expertise is sufficient to perform the categorization and that the results of the different evaluations (PRA and non-PRA) are used in an integrated, systematic process as required by 10 CFR 50.69(c)(1)(ii). Based on the above, the NRC staff finds that the licensee's proposed IDP would have the expertise to meet the requirements in 10 CFR 50.69(c)(2) and that the additional descriptions of the IDP characteristics, training, processes, and decision guidelines are consistent with NEI 00-04, as endorsed by the NRC in RG 1.201.

Conclusion for Key Principle 4 The NRC staff reviewed the acceptability of the licensees internal events PRA (including internal floods), fire PRA, and seismic PRA. The staff also reviewed the use of PRA importance measures and integrated importance measures, the use of non-PRA methods, risk sensitivity studies, and integrated decision-making. Based on these reviews and the findings described above, the NRC staff has determined that the proposed change satisfies the fourth key principle for risk-informed decision-making identified in RG 1.174, Revision 3 and ensures that any potential increases in risk are small as required by 10 CFR 50.69(c)(1)(iv).

3.3.5 Key Principle 5: Monitor the Impact of the Proposed Change NEI 00-04 provides guidance that includes programmatic configuration control and a periodic review to ensure that all aspects of the 10 CFR 50.69 program (including traditional engineering analyses) and PRA models used to perform the risk assessment continue to reflect the as-built and as-operated plant and that plant modifications and updates to the PRA are continually incorporated.

Program Documentation and Change Control Paragraph 50.69(f) of 10 CFR requires program documentation, change control, and records. In Section 3.2.6, PRA Maintenance and Updates, of the LAR dated May 28, 2024, the licensee stated that it will implement a process that addresses the requirements in Section 11 of NEI 00-04, pertaining to program documentation and change control records. In Section 3.1.1 of the LAR enclosure, the licensee states that the RISC categorization process documentation will include the following 10 elements:

1. Program procedures used in the categorization

2. System functions, identified and categorized with the associated bases

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members The NRC staff also recognizes that for facilities licensed under 10 CFR Part 50, Appendix B Criterion VI, Document Control, procedures are considered formal plant documents requiring that [m]easures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality.

The elements provided in Section 3.1.1 of the LAR enclosure, in addition to the list of items provided in Attachment 1 of enclosure 2 to the LAR, will ensure that Dresdens 10 CFR 50.69 categorization process will be documented in formal licensee procedures. This is consistent with Section 11 of NEI 00-04, as endorsed by the NRC in RG 1.201, Revision 1. Based on the above, the NRC staff finds that the procedures will be sufficient for meeting the 10 CFR 50.69(f) requirement for program documentation, change control, and records.

Periodic Review (NEI 00-04, Section 12)

Paragraph 50.69(e) of 10 CFR requires that periodic updates to the licensees PRA and SSC categorization must be performed. Changes over time to the PRA and to the SSC reliabilities are inevitable, and such changes are recognized by the 10 CFR 50.69(e) requirement for periodic updates.

In Section 3.2.6, PRA Maintenance and Updates, of the LAR dated May 28, 2024, the licensee described the process for maintaining and updating the DNPS PRA models used for the 10 CFR 50.69 categorization process. Consistent with NEI 00-04, the licensee confirmed that the DNPS risk management process ensures that the PRA models used in this application continue to reflect the as-built and as-operated plant. The licensees process includes provisions for: monitoring issues affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, and industry operational experience); assessing the risk impact of unincorporated changes; and controlling the model and associated computer files.

The process also includes reevaluating previously categorized systems to ensure the continued validity of the categorization. Routine PRA updates are performed every two refueling outages at a minimum.

Section 12.1 of NEI 00-04, Revision 0, states, in part, [s]cheduled periodic reviews (e.g. once per two fuel cycles in a unit) should evaluate new insight resulting from available risk information. In Section 3.5, of the LAR enclosure, the licensee states, in part:

Scheduled periodic reviews no longer 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.

The NRC staff finds that the risk management process described by the licensee in the LAR is consistent with the guidance in Section 12 of NEI 00-04. Considering the above, the staff has determined that the proposed change satisfies the fifth key principle for risk-informed decision-making identified in RG 1.174.

4.0 CHANGES TO THE OPERATING LICENSE Based on the NRC staffs review of the LAR, as supplemented, the staff identified specific actions, as described below, that are necessary to support the NRC staffs conclusion that the proposed program meets the requirements in 10 CFR 50.69 and the guidance in NEI 00-04, as endorsed in RG 1.201, Revision 1. Note: Additional actions (e.g., final procedures and proposed alternative treatment) need not, and have not been submitted, or reviewed by the NRC staff for issuance of the SE but will be completed before implementation of the program as specified in the 10 CFR 50.69 rule.

The NRC staffs finding on the acceptability of the PRA evaluation in the licensees proposed 10 CFR 50.69 process is conditioned upon the license condition provided below.

The licensee proposed the following amendment to the RFOLs for DNPS, Units 2 and 3. The proposed license condition states:

CEG is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire, seismic events; 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 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except external flooding and extreme winds and tornadoes; an external flood safe shutdown equipment list for external floods; and a high wind safe shutdown equipment list for extreme winds and tornadoes.

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 probabilistic risk assessment approach to seismic margins approach).

The NRC staff finds that the proposed license condition is acceptable because it adequately implements 10 CFR 50.69 using models, methods, and approaches consistent with the applicable NRC and NRC-endorsed guidance.

The NRC staff notes that the guidance for implementing 10 CFR 50.69 provided by the Commission in the Federal Register notice dated November 22, 2004 (69 FR 68008),

Section III.4.10.2, Section 50.36 Technical Specifications, stated that the 10 CFR 50.69 rule does not include 10 CFR 50.36 in the list of special treatment requirements that may be replaced by the alternative 10 CFR 50.69 requirements for RISC-3 and RISC-4 SSCs when implementing a 10 CFR 50.69 license amendment. As a result, the NRC staff does not consider the TSs (including Improved Technical Specifications (ITS) and the associated Technical Requirements Manual) to be part of the 10 CFR 50.69 rule. Therefore, the licensee needs to address proposed changes to its TS separately.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendment on July 17, 2025. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types of effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, as published in Federal Register on August 6, 2024 (89 FR 63991), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

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

8.0 REFERENCES

[1] Humphrey, M. D., Constellation Energy Generation, LLC, to U.S. Nuclear Regulatory Commission, "Application to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants",

May 28, 2024, Agencywide Documents Access and management System (ADAMS)

Accession No. ML24149A261).

[2] Humphrey, M. D., Constellation Energy Generation, LLC to U.S. Nuclear Regulatory Commission, "Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b,'... and 10 CFR 50.69...",

March 21, 2025 (ML25080A153).

[3] Humphrey, M. D., Constellation Energy Generation, LLC to U.S. Nuclear Regulatory Commission, "Response to the Request for Additional Information Related to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSF...", April 23, 2025 (ML25113A134).

[4] Arora, Surinder, NRC letter to Rhoades, David, Constellation Energy Generation, Plan for the Regulatory Audit in support of License Amendment Requests to Adopt TSTF-505, Revision 2, TSTF-591, Revision 0, and 10 CFR 50.69" dated July 24, 2024 (ML24204A186).

[5] Arora, Surinder, NRC letter to Constellation Energy Generation, Request for Additional Information, by the Office of Nuclear Reactor Regulation, Dresden 2 and 3 - License Amendment to Adopt TSTF-505 and TSTF-591, Constellation Energy Generation, LLC",

March 25, 2025 (ML25084A129).

[6] Arora, Surinder, NRC letter dated May 20, 2025, to Rhoades, David, Constellation Energy Generation, Summary Report for the Regulatory Audit in support of License Amendment Requests to Adopt TSTF-505, Revision 2, TSTF-591, Revision 0, and 10 CFR 50.69" May 20, 2025 (ML25098A179).

[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 (ML17317A256).

[8] NRC, Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed Activities," Revision 2, March 2009 (ML090410014).

[9] NRC, Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020 (ML20238B871).

[10] 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 (ML061090627).

[11] NRC, NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," Revision 1, March 2017 (ML17062A466).

[12] NRC, NUREG-0800, Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance,"

Revision 0, June 2007 (ML071700658).

[13] NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, July 2005 (ML052910035).

[14] Anderson, V.K., NEI, letter to Stacey Rosenberg, U.S. Nuclear Regulatory Commission, "Final Revision of Appendix X to NEI 05-04/07-12/12-[13], Close-Out of Facts and Observations (F&Os)", February 21, 2017 (package, ML17086A431).

[15] Giitter, J., and Ross-Lee, M.J., NRC, letter to Greg Krueger, NEI, "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os)", May 3, 2017 (ML17079A427).

[16] Humphrey, M. D., Constellation Energy Generation, LLC to NRC, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b,' and TSTF-591...", May 8, 2024 (ML24129A135).

[17] ASME and ANS PRA standard ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications", February 2009, New York, NY (Copyright).

[18] Arora, Surinder, NRC letter to Rhoades, David, Constellation Energy Generation, Dresden Nuclear Power Station, Units 2 and 3 - Issuance of Amendment Nos. 286 and 279 Re:

Adoption of TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4B and TSTF-591, Revise Risk Informed Completion Time (RICT) Program (EPID L-2024-LLA-0061), dated July 30, 2025 (ML25196A299).

[19] ASME and ANS, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, PRA Standard ASME/ANS RA-Sb-2013, New York, NY, September 2013.

[20] NEI 12-13, "External Hazards PRA Peer Review Process Guidelines", August 2012, (ML12240A027).

[21] Franovich, M., NRC, "U.S. Nuclear Regulatory Commission Acceptance of Nuclear Energy Institute (NEI) Guidance NEI 12-13, 'External Hazards PRA Peer Review Process Guidelines'", March 7, 2018 (ML18025C025).

[22] NRC, Generic Letter 88-20, Supplement 4, "Individual Plant Examination of External Events (IPEEEs) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), June 28, 1991 (ML031150485).

[23] Nuclear Management and Resources Council, Inc., NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management", December 1991 (ML14365A203).

[24] Markley, Michael, NRC, letter to Vice President, Operations, Arkansas Nuclear One, Entergy Operations, Inc., Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO-2 R&R-004, Revision 1. Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 & 3 Moderate and High Energy Systems, April 22, 2009 (ML090930246).

[25] NRC, Regulatory Issue Summary 2015-06, "Tornado Missile Protection", June 2015 (ML15020A419).

[26] Dresden Nuclear Power Station, Unit Nos. 2 and 3, (UFSAR), Revision 15, June 2023.

[27] U.S. Atomic Energy Commission, Regulatory Guide 1.78, "Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release", June 1974 (ML003740298).

[28] ASME, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities, ASME Code Case, N-660", July 2002.

Principal Contributors: Adrienne Brown, NRR Todd Hilsmeier, NRR Michael Swim, NRR Keith Tetter, NRR Date of Issuance: August 7, 2025

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