ML23320A306

Issuance of Amendment Nos. 298 and 291 Regarding Revising the Fire Protection Program in Support of Reactor Coolant Pump Seal Replacement Project

ML23320A306
Person / Time
Site:	Turkey Point
Issue date:	01/22/2024
From:	Michael Mahoney NRC/NRR/DORL/LPL2-2
To:	Coffey B Florida Power & Light Co, NextEra Energy Seabrook
Mike Mahoney, NRR/DORL/LPL2-2
References
EPID: L-2022-LLA-0128
Download: ML23320A306 (25)

Text

January 22, 2024

Mr. Bob Coffey Executive Vice President, Nuclear and Chief Nuclear Officer Florida Power & Light Company Mail Stop: EX/JB 700 Universe Blvd.

Juno Beach, FL 33408

SUBJECT:

TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 - ISSUANCE OF AMENDMENT NOS. 298 AND 291 REGARDING REVISING THE FIRE PROTECTION PROGRAM IN SUPPORT OF REACTOR COOLANT PUMP SEAL REPLACEMENT PROJECT (EPID: L-2022-LLA-0128)

Dear Mr. Coffey:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 298 to Subsequent Renewed Facility Operating License No. DPR-31 and Amendment No. 291 to Subsequent Renewed Facility Operating License No. DPR-41 for the Turkey Point Nuclear Generating Unit Nos. 3 and 4, respectively. These amendments revise the Fire Protection Program (FPP) in support of reactor coolant pump (RCP) seal replacements in response to your application dated August 26, 2022, as supplemented by letters dated May 31, July 27, August 25, and October 31, 2023.

Specifically, the amendments are requested in accordance with the units operating licenses, paragraph 3.D, Fire Protection, for FPP changes that may be made without prior NRC approval. One of the criteria for such a change is that the risk increase resulting from the change is less than 1 x 10-7/year (yr) for core damage frequency (CDF) and less than 1 x 10-8/yr for large early release frequency (LERF). The change is to support replacement of the currently installed RCP seals with the Framatome RCP hydrostatic seal package equipped with the passive shutdown seal.

B. Coffey

A copy of the related safety evaluation is also enclosed. Notice of issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Michael Mahoney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket Nos. 50-250 and 50-251

Enclosures:

1. Amendment No. 298 to DPR-31

2. Amendment No. 291 to DPR-41

3. Safety Evaluation

cc: Listserv FLORIDA POWER & LIGHT COMPANY

DOCKET NO. 50-250

TURKEY POINT NUCLEAR GENERATING UNIT NO. 3

AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE

Amendment No. 298 Subsequent Renewed License No. DPR-31

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

A. The application for amendment by Florida Power & Light Company (the licensee) dated August 26, 2022, as supplemented by letters dated May 31, July 27, August 25, and October 31, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, by Amendment No. 298, the licensee is authorized to make changes to the fire protection program, as set forth in the application dated August 26, 2022, as supplemented by letters dated May 31, July 27, August 25, and October 31, 2023. The licensee shall make changes as described in the licensees application and supplemental letters and as evaluated in t he NRC staffs safety evaluation issued with this amendment.

3. This license amendment is effective as of the date of its issuance and shall be implemented by the completion of refueling outage 3R35 (spring 2026) for Unit No. 3.

FOR THE NUCLEAR REGULATORY COMMISSION

David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Subsequent Renewed Facility Operating License and Technical Specifications

Date of Issuance: January 22, 2024 FLORIDA POWER & LIGHT COMPANY

DOCKET NO. 50-251

TURKEY POINT NUCLEAR GENERATING UNIT NO. 4

AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE

Amendment No. 291 Subsequent Renewed License No. DPR-41

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

A. The application for amendment by Florida Power & Light Company (the licensee) dated August 26, 2022, as supplemented by letters dated May 31, July 27, August 25, and October 31, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, by Amendment No. 291, the licensee is authorized to make changes to the fire protection program, as set forth in the application dated August 26, 2022, as supplemented by letters dated May 31, July 27, August 25, and October 31, 2023. The licensee shall make changes as described in the licensees application and supplemental letters and as evaluated in t he NRC staffs safety evaluation issued with this amendment.

3. This license amendment is effective as of the date of its issuance and shall be implemented by the completion of refueling outage 4R33 (spring 2025) for Unit No. 4.

FOR THE NUCLEAR REGULATORY COMMISSION

David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Subsequent Renewed Facility Operating License and Technical Specifications

Date of Issuance: January 22, 2024 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NOS. 298 AND 291

TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NOS. DPR-31 AND DPR-41

FLORIDA POWER & LIGHT COMPANY

TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4

DOCKET NOS. 50-250 AND 50-251

1.0 INTRODUCTION

By letter dated August 26, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22243A161), as supplem ented by letters dated May 31, July 27, August 25, and October 31, 2023 (ML23151A607, ML23208A191, ML23240A153, and ML23304A156, respectively), Florida Power and Light Company (FPL, the licensee) submitted a license amendment request (LAR) to revise the Fire Protection Program (FPP) in support of reactor coolant pump (RCP) seal replacements at Turkey Point Nuclear Generating (Turkey Point) Unit Nos. 3 and 4.

Specifically, the amendments were requested in accordance with the units operating licenses, paragraph 3.D, Fire Protection, for FPP changes that may be made without prior U.S. Nuclear Regulatory Commission (NRC) approval. One of t he criteria for such a change is that the risk increase resulting from the change is less than 1 x 10-7/year (yr) for core damage frequency (CDF) and less than 1 x 10-8/yr for large early release frequency (LERF). The proposed change is to support replacement of the currently installed RCP seals with the Framatome RCP hydrostatic seal package equipped with the passive shutdown seal (PSDS).

The supplements dated May 31, July 27, August 25, and October 31, 2023, provided additional information that clarified and corrected the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 27, 2022 (87 FR 79359).

Enclosure 3

2.0 REGULATORY EVALUATION

2.1 Regulatory Basis

2.1.1 Regulatory Requirements

Title 10, Energy, of the Code of Federal Regulations (10 CFR) section 50.48, Fire protection, provides the requirements for nuclear power plant fire protection.

The regulations in 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, referring to National Fire Protection Association (NFPA) Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition, include provisions for licensees to make changes to their approved FPPs once the transition to a 10 CFR 50.48(c) license is complete, as described in the transitioned plants license condition.

The regulations in 10 CFR 50.55a, Codes and standards, incorporate by reference the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section III.

The regulations in 10 CFR 50.59, Changes, tests, and experiments, specify the requirements by which licensees, under certain conditions, may make changes to their facilities and procedures as described in the final safety analysis report (FSAR) (as updated), also referred to as the updated final safety analysis report (UFSAR), and conduct tests or experiments not described in the FSAR (as updated) without obtaining a license amendment pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit.

2.1.2 Regulatory Guidance

NRC Regulatory Guide (RG) 1.205, Revision 2, Risk-informed, Performance-based Fire Protection for Existing Light-water Nuclear Power Plants (ML21048A448), includes provisions for licensees to make changes to their approved FPPs once the transition to a 10 CFR 50.48(c) license is complete. With any changes, the licensee is required to perform a plant change evaluation to demonstrate the acceptability of the change in terms of risk increase, defense-in-depth, and safety margin.

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-specific Changes to the Licensing Basis (ML17317A256), provides acceptance guidance applicable to NFPA 805 plant change evaluations.

NUREG-1852, Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire, dated October 2007 (ML073020676), provides guidance for assessing the adequacy of time margins associated with the completion of manual operator actions.

NUREG-2178, Refining and Characterizing Heat Re lease Rates from Electrical Enclosures During Fire, Volume 2, Fire Modeling Guidance for Electrical Cabinets, Electric Motors, Indoor Dry Transformers, and the Main Control Board (ML20168A655), provides guidance for improved methods that will increase the realism in the modeling of selected ignition sources.

NUREG-2230, Methodology for Modeling Fire Grow th and Suppression Response for Electrical Cabinet Fires in Nuclear Power Plants (ML20157A148), provides guidance for obtaining a more realistic estimate of plant risk based on fire probabilistic risk assessments (FPRAs).

NUREG-2233, Methodology for Modeling Transient Fires in Nuclear Power Plant Fire Probabilistic Risk Assessment (ML20289A568), provides guidance for probabilistic distributions and detailed fire modeling input parameters for the modeling of transient fires in FPRAs.

For circumstances involving both changes result ing from plant modifications and the evaluation of important human actions, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (the SRP),

Chapter 18, Human Factors Engineering (ML16125A114), provides guidance regarding the use of NUREG-1764, Guidance for the Review of Changes to Human Actions, Revision 1 (ML072640413). NUREG-1764 provides guidance for reviewing changes in human actions, such as those that are credited in nuclear power plant safety analyses.

In conducting reviews of the human factors engineering (HFE) aspects of licensing submittals for LWR facilities, the NRC staff applies the gu idance in the SRP. Chapter 18 of the SRP provides guidance for the review of HFE considerations of plant modifications and important human actions. SRP Chapter 18, section IV, Review Procedures, states, in part, that [t]he degree to which the NRC staff applies the review methodology in this SRP will reflect the specific circumstances of individual applications.

Sections 7.2.1, Design Bases, Core Protection Systems, and 7.5.1, Design Bases, Engineered Safety Features Protection Systems, of the Turkey Point UFSAR (ML22303A017) include a commitment that facility design will conform to Proposed General Design Criterion (GDC) 11, Control Room, as origina lly published in 1967 (ML003674723). Proposed GDC 11 states, in part, that the facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled.

RG 1.187, Revision 3, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments (ML21109A002), provides licensees with a method that the NRC staff considers acceptable for use in complying with the NRCs regulations on the process by which licensees, under certain conditions, may make changes to their facilities and procedures as described in the UFSAR and conduct tests or experiments not described in the UFSAR without obtaining a license amendment pursuant to 10 CFR 50.90.

2.2 Description of the Proposed Changes

The licensee proposed changes to the Turkey Point FPP due to the planned replacement of the currently installed RCP seals with the Framatome RCP hydrostatic seal package equipped with the PSDS (Framatome PSDS). The licensee stated that the replacement Framatome PSDSs have been evaluated for installation at Turkey Point under the criteria of 10 CFR 50.59 to determine if any attribute requires prior NRC approval. The licensee stated that no attribute for the installation of the replacement seals was identified as requiring prior NRC approval.

However, the approved Turkey Point Subsequent Renewed Facility Operating License (SRFOL) fire protection license conditions (i.e., Paragraph 3.D) provide that plant changes which may be made without prior NRC approval must meet certain criteria.

One of the criteria for such a change is that the risk increase resulting from the change is less than 1 x 10-7/yr for CDF and less than 1 x 10-8/yr for LERF. The proposed change must also be

consistent with the defense-in-depth philosophy and must maintain sufficient safety margins.

The licensee performed a Fire probabilistic risk assessment (PRA) evaluation for the replacement of the currently installed RCP seals with the Framatome PSDS and concluded that the risk threshold for self-approval is exceeded. Therefore, the licensee submitted the LAR for prior NRC approval for the Fire PRA treatment of RCP seal leakage for the new seal design.

The LAR would update the Turkey Point Fire PRA to incorporate the Framatome PSDS seal model in conjunction with model refinements to ac hieve a more realistic assessment of the plant risk and the risks associated with this planned modification. These refinements include the modeling of interruptible fires as defined in NUREG-2230 and updated transient heat release rates as defined in NUREG-2233, as well as guidance provided in NUREG-2178 regarding a more realistic assessment of the impact of fires near walls and corners.

2.3 Background

In its application dated August 26, 2022, the licensee provided the following system description:

Westinghouse Model 93 RCPs are installed at Turkey Point.

An RCP seal prevents reactor coolant from exiting the RCP body along the pumps impeller shaft. Turkey Point currently has a three-stage RCP seal design installed on Unit 3 and Unit 4. These seal types were first installed during the Unit 3 fall 2015 and Unit 4 spring 2016 refueling outages. The current seal has three identical hydrodynamic mechanical stages in series, plus an abeyance seal designed to be installed on the Westinghouse RCPs.

Prior to installation of the current RCP seals, AREVA RCP seals were installed.

The replacement of the existing AREVA RCP seals with the current RCP seals was necessary to reduce seal leakage to allow the Turkey Point Units to cope with a beyond design basis event with an extended loss of AC [alternating current] power (ELAP).

Turkey Point also transitioned to NFPA 805 as the basis for its Fire Protection Program in 2015. Similar to the effect on the beyond design basis event, a reduction in seal leakage increases the operator response time for fire damage scenarios and reduces the risk evaluated in the Fire PRA.

Approval of Turkey Points transition to NFPA 805 for the Fire Protection Program was documented in Turkey Point Amendments 262 and 257

[ML15061A237]. Included in this approval is the use of a PRA model specific to the current RCP seal design. This implementation item was changed in 2019.

In Turkey Point Amendments 286 and 280 [ML19064A903], the NRC approved a revision to use the guidance from WCAP-16175-P-A [Model for Failure of RCP Seals Given Loss of Seal Cooling in CE [Combustion Engineering] NSSS

[Nuclear Steam Supply System] Plants, ML071130383] to model the RCP seals.

The Fire PRA was updated with selected data, specifically, the Human Error Probability (HEP) values associated with reduced time to trip the RCPs on loss of seal cooling. The time to trip the RCPs following a loss of seal cooling was established as 20 minutes from WCAP-16175-P-A, and the RCP seal failure

probabilities in the PRA model were adjusted to be consistent with those presented in WCAP-16175-P-A appropriate for the current RCP seals.

Additionally, in its application dated August 26, 2022, the licensee provided the following description of the Framatome RCP Hydrostatic Seal with PSDS:

The Framatome RCP hydrostatic seals have three seal stages, and the Framatome seal package includes a fourth seal that is referred to as the PSDS.

Leakage from the RCPs along the RCP shaft is controlled by a seal package that consists of three mechanical shaft seals arranged in series and referred to as seal No. 1, No. 2 and No. 3. During normal operation, high pressure seal water is injected (~8 gpm [gallons per minute]) by the Chemical and Volume Control System into the seal chamber to cool the seals and prevent the leakage of high temperature reactor coolant into the seal chamber. The injection water, which overcomes the normal RCS [reactor coolant system] operating pressure, enters the pump through a connection on the thermal barrier flange.

The PSDS was specifically designed to improve the ability of the RCP seal package to control RCS leakage during ELAP and station blackout (SBO) events, when the seal cooling normally provided by the thermal barrier and the seal injection flow is unavailable for an extended period of time. After seal cooling is lost, the fluid temperature upstream of the shaft seals will rise rapidly and the leakage of reactor coolant through the RCP seals will increase. The PSDS is a fourth seal within the Framatome RCP hydrostatic seal package that remains inactive during normal operation of the RCP. It actuates automatically in response to the temperature rise that takes place within the seal assembly which results from a loss of cooling to the RCP seals. After the PSDS has actuated, it seals the reactor coolant leak path that exists between the pump shaft and the lower seal housing.

The PSDS is a passively actuated mechanical seal that is designed to provide very low leakage through the installed RCP seals in the event of an ELAP. The PSDS is pre-assembled into the No.1 seal insert. The PSDS will remain in the inactive state until an event causes the PSDS to be exposed to temperatures above its actuation temperature.

3.0 TECHNICAL EVALUATION

3.1 Deterministic Review

3.1.1 Mechanical Review

The NRC staff reviewed the mechanical engineering aspects of the proposed new RCP seal design at Turkey Point. In particular, the NRC staff reviewed the LAR, as supplemented, performed a detailed audit of information made available by the licensee in its electronic reading room, and conducted several telephone conferences with licensee representatives and their contractors. An audit summary was issued on September 25, 2023 (ML23249A241). In this section, the NRC staff summarizes its mechanical review and provides its findings regarding the proposed new RCP seal design by Framatome for Turkey Point.

3.1.1.1 RCP Seal Current Design

Westinghouse Model 93 RCPs are currently installed at Turkey Point Unit Nos. 3 and 4. The RCP seal prevents nuclear reactor coolant from exiting the RCP body along the pumps impeller shaft. The licensee installed the current three-stage Flowserve RCP seals for Turkey Point Unit Nos. 3 and 4 during the fall 2015 and spring 2016 refueling outages, respectively. These RCP seals include a passive back-up sealing device, referred to as the Abeyance Seal, which activates upon the complete failure of the three primary seal stages.

3.1.1.2 RCP Seal Proposed Design Modification

In the LAR, the licensee discussed the proposed replacement of the currently installed RCP seals with a three-stage Framatome RCP hydrostatic seal package equipped with the PSDS.

The licensee stated that the replacement Framatome RCP hydrostatic seal package with the PSDS has been evaluated for installation in Turkey Point under the criteria of 10 CFR 50.59.

The licensee stated that no attribute for the installation of the replacement seals has been identified as requiring prior NRC approval. Therefore, this LAR only addresses the change in the Fire PRA supporting the FPP due to the planned installation of the Framatome RCP hydrostatic seal with the PSDS. The licensee stated that operating procedures and operator training will be reviewed for consistency with the conditions and timing evaluated in site-specific analyses and will be revised, as necessary, to support the planned installation of the Framatome RCP hydrostatic seals with the PSDS. The licensee stated that any required procedure changes needed as part of the planned installation of the Framatome RCP hydrostatic seal with the PSDS will be evaluated under 10 CFR 50.59.

3.1.1.3 Framatome RCP Seal Model and Additional Plant Modifications

In its letter dated August 25, 2023, the licensee stated that the PSDS manufacturer has not assigned a model number to the PSDS that is planned to be installed at Turkey Point. The PSDS supplier has assigned the PSDS with Part Number 1251149-151. The licensee also stated that plant modifications to support the planned installation of the Framatome RCP seals with the PSDS would include the configuration of the plant piping to facilitate the required process connections to the RCP seal (i.e., seal leakoff piping and seal bypass). The modifications would also include the installation of instrumentation needed to monitor seal performance and to detect abnormal seal operating conditions (i.e., temperature, pressure, and leakoff flow).

3.1.1.4 RCP Seal Proposed Design and ASME BPV Code to the LAR and Framatome document 152-9348792-000, Reactor Coolant Pump Seal and Seal Housing Procurement Specification for Turkey Point Units 3 and 4, listed in Section 5.0 of the audit summary dated September 25, 2023, provide that the reactor coolant pressure boundary (RCPB) components of the RCP seal assembly are required to be designed, manufactured, and tested in accordance with the ASME BPV Code,Section III, Class 1, 2010 Edition with 2011 Addenda. In the Framatome RCP seal assembly, the lower seal housing is the only component that is classified as part of the RCPB and, as such, will be furnished to meet the ASME BPV Code requirements. Therefore, the planned installation of the Framatome RCP seals will not have an adverse impact on safety margin as they will meet the requirements of the ASME BPV Code,Section III, Class 1, 2010 Edition with the 2011 Addenda, as incorporated by reference in 10 CFR 50.55a.

3.1.1.5 PSDS Design

As described in the LAR, the PSDS is a passively actuated mechanical seal that is designed to provide very low leakage through the RCP seals in the event of an ELAP. The PSDS is installed as an integral portion of the existing No. 1 RCP seal system. The PSDS will remain in the inactive state until an event causes the PSDS to be exposed to temperatures above its actuation temperature.

3.1.1.6 Framatome RCP Seal Actuation Temperature

According to the LAR, actuation of the PSDS occurs when the coolant (leak-off) temperature exceeds the PSDS actuation temperature. Upon actuation, the PSDS will form a seal around the shaft. The NRC staff did not identify any concer ns with the RCP seal actuation temperature.

3.1.1.7 Potential Risk Increase due to Proposed New RCP Seals

As part of its review, the NRC staff considered the potential for an increase in risk as a result of the planned installation of the new RCP seal design at Turkey Point. In particular, the NRC staff considered the design of the new Framatome RCP seals relative to the design of the currently installed RCP seals, the reduction in the available operator action time to trip the RCPs, and the modelling of the potential failure modes. To address these considerations, plant operations personnel at Turkey Point will monitor and trend t he RCP seal performance parameters from the main control room, including: (a) seal water injection flow and temperature; (b) No. 1 seal differential pressure, leak-off flow, and leak-off temperature; (c) No. 2 seal leak-off flow; and (d) No. 3 seal leak-off flow. These actions will enable the licensee to identify any inadvertent seal actuation and any degradation issues which may affect seal performance. Therefore, the NRC staff does not consider the new RCP seal planned installation as a potential increase in risk during Turkey Point operations.

3.1.1.8 PSDS Actuation and Operator Action

As part of its review, the NRC staff evaluated the PSDS actuation and operator actions described in the licensee submittals and in proprietary information made available by the licensee. At Turkey Point, the PSDS will actuate when reaching a certain temperature. During normal plant operation, the temperature of the PSDS is approximately equal to the temperature of the RCP seal injection water. If seal cooling is lost, reactor coolant from the RCS cold leg will flow upward through the RCP, past the thermal barrier, into the free volume of the RCP main flange assembly and No. 1 seal cavity, and past the No. 1 seal and the PSDS. The time required for the reactor coolant from the RCS cold leg to reach the PSDS is the time required for the water originally in the No. 1 seal cavity to be purged past the No. 1 seal by the in-flowing reactor coolant. This purge time is directly proportional to the No. 1 seal leak rate.

According to the licensees letter dated October 31, 2023, the operator actions in response to a loss of seal cooling (LOSC) event would be to manually trip the reactor, manually trip the RCPs, stabilize plant conditions, prepare for RCS cooldown via natural circulation, and commence natural circulation RCS cooldown. The operator response requires numerous procedural steps to be completed using various procedures. Depending on the actual plant conditions and the procedural flow path, the time required before reaching the procedural step for initiating a controlled RCS natural circulation cooldown would be 30 to 45 minutes. During the controlled

natural circulation cooldown of the RCS, the RCS temperature would be gradually reduced at a rate within technical specification limits; generally, at a rate of approximately 50°F per hour.

While the operator response is taking place, the hot reactor coolant from the RCS cold leg will begin to flow upward through the RCP, into the RCP seal housing, and through the No. 1 seal.

The reactor coolant will flash after flowing through the No. 1 seal, but its temperature will remain above the actuation temperature of the PSDS. The time required for reactor coolant to reach the downstream side of the No. 1 seal and actuate the PSDS is dependent upon the initial No. 1 seal leak rate. For example, a low initial leak rate will extend the PSDS actuation time while a high initial leak rate will reduce the PSDS actuation time. Even if the No. 1 seals were operating at their lowest leakage limit, the PSDS would actuate prior to any significant reduction in the RCS cold leg temperature from the commencement of the controlled natural circulation cooldown. Therefore, the operator actions taken to initiate a controlled natural circulation cooldown will not affect PSDS actuation.

Operator actions other than commencing a controlled RCS natural circulation cooldown can take place while recovering from an LOSC event that might have an adverse effect on a PSDS that has already actuated (e.g., the restart of an RCP). To account for this possibility, the licensee stated in its letter dated October 31, 2023, that the applicable Turkey Point procedures will be revised to require checking the status of PSDS actuation prior to taking such actions.

3.1.1.9 Framatome RCP Seal and Operational Aspects

Based on operating experience, the currently installed RCP seals at Turkey Point are performing satisfactorily. The Framatome RCP No. 1 and No. 2 seals are similar to the Westinghouse No. 1 and No. 2 seal design, while the No. 3 seal is a Framatome proprietary balanced nose design. The only substantive difference between the currently installed RCP seals and the Framatome RCP seals would be the addition of the PSDS. The Framatome RCP seals are an industry-accepted design installed in multiple nuclear power plants. The Framatome RCP seals are intended to provide improved performance through multiple operating cycles at Turkey Point.

3.1.1.10 Licensees Justification for Reduction in Time to Trip RCPs

In its letter dated August 25, 2023, the licensee justified the reduction in time to trip the RCPs during an LOSC event from 20 minutes to 16 minutes. The licensee stated that the currently installed RCP seals are assumed to continue to perform their function if the RCP trips and if the controlled bleed-off is isolated within 20 minutes of an LOSC event. For the proposed replacement Framatome RCP seals, an available 16-minute operator action time to trip the RCPs during an LOSC event is established to ensure that the PSDS will reliably seal the leak path between the RCP shaft and RCP lower housing. For the PSDS to establish a reliable seal, the RCPs must be manually tripped, and the RCP coastdown must be complete prior to PSDS actuation. Therefore, the operator response time plus the RCP coastdown time must be less than the PSDS actuation time.

The licensee stated that the operator action time to trip RCPs following an LOSC event is a proceduralized time-sensitive action related to a fire response. At Turkey Point, operators are trained to ensure that this action can be completed in the assumed time frame. The licensee reported that the validation time for the operator response to trip the RCPs following an LOSC

event in 6.17 minutes. This time is much less than the 16 minutes assumed in the Fire PRA evaluation. Therefore, the NRC staff did not identify concerns with the operator response time to trip the RCPs during an LOSC event.

3.1.1.11 Mechanical Review Conclusion

Based on its review of the mechanical engineering aspects of the proposed new RCP seal design at Turkey Point, the NRC staff finds that:

a. The licensee evaluated the proposed replacement Framatome RCP hydrostatic seal package with the PSDS for installation at Turkey Point under the criteria of 10 CFR 50.59.

b. The Framatome RCP seals are an industry-accepted design installed in several nuclear power plants and will provide improved performance through multiple operating cycles.

c. The PSDS is a passively actuated mechanical seal that is designed to provide very low leakage through RCP seals in the event of an ELAP.

d. The installation of the Framatome RCP seals will not have an adverse impact on safety margin as they meet the requirements of the ASME BPV Code Section III, Class 1, 2010 Edition with 2011 Addenda, as incorporated by reference in 10 CFR 50.55a.

e. Actuation of the PSDS occurs when the coolant (leak-off) temperature exceeds the PSDS actuation temperature. The licensee has justified that actuation of the PSDS will occur in a timely manner.

f. Plant modifications to be performed to support the planned installation of the Framatome seals with the PSDSs are limited to the configuration of the plant piping to facilitate the required process connections to the seals and the installation of the instrumentation required to monitor seal performance and detect abnormal seal operating conditions.

g. The licensee justified the 16-minute operator action time to trip the RCPs during an LOSC event to ensure that the PSDS will reliably seal the leak path between the RCP shaft and RCP lower housing.

Based upon its mechanical review of the information provided in the LAR, as supplemented, and responses to audit questions, the NRC staff concludes that the licensee has justified the proposed use of the Framatome RCP seal design at Turkey Point.

3.1.2 Human Factors Engineering Review

The NRC staff reviewed the HFE-related aspects of the LAR using the guidance of NUREG-1764.

3.1.2.1 Description and Assessment of the Credited Operator Actions

In the LAR, FPL described the operator actions needed to support the proposed replacement of the currently installed RCP seals with the Framatome RCP hydrostatic seal package equipped with the PSDS. Specifically, the LAR states that the LOSC event must consider the time available for the operator to manually trip the RCPs. The current time to trip the RCPs following

an LOSC was established at 20 minutes. The proposed change in the RCP seal design would reduce the time available to verify that the RCPs are tripped in the event of a complete loss of RCP seal cooling to 16 minutes.

In accordance with the guidance provided in Chapter 18 of the SRP, the NRC staff used a graded approach to evaluate the HFE considerations related to the changes described in the LAR. Because the licensee submitted a non-risk-informed LAR, the NRC staff used a qualitative approach under NUREG-1764 in determining the risk significance of the proposed change and the corresponding level of review. In accordance with the generic risk categories established in Appendix A, Generic Human Actions that are Risk-important, to NUREG-1764, the operator actions associated with RCP seal cooling are classified as risk-important human actions. RCP trip on an LOSC is listed on Table A.2, Generic PWR [pressurized-water reactor] Human Actions That Are Risk-Important, which is the Group 1 risk-important qualitative list of items for PWRs. The initial qualitative screening per NUREG-1764, section 2.4.3.2, Generic HA

[human action] Method (Method 2), would be Level I, but was reduced to the Level II review based upon the factors described in NUREG-1764, section 2.3.5.2, Adjusting the Level of Human Factors Engineering Review.

In conducting this qualitative assessment, key areas of consideration included the following:

whether the requested change would introduce new human actions.

whether the requested change would give personnel a new functional responsibility that differs from their normal responsibilities.

whether the requested change would significantly modify the way in which personnel perform their tasks.

whether the requested changed change would create a new context for task performance; and whether the requested change would significantly change any of the following:

o the human-system interfaces used by personnel to perform the task.

o the procedures that personnel use to perform the task; or o the training associated with the task.

Based upon the information provided in the LAR, as supplemented, as detailed later in this evaluation, the NRC staff determined that significant changes were not expected to occur within these areas. From this qualitative assessment, the NRC staff then concluded that the review could be reduced to the Level II review. Using the screening and review guidance of NUREG-1764, the NRC staff subsequently determined that the appropriate Level II review criteria included deterministic, analysis, design, and human action verification criteria. The relevant Level II review criteria are considered in the subsequent portions of this evaluation.

3.1.2.2 General Deterministic Review Criteria

Section 4.1, General Deterministic Review Cr iteria, of NUREG-1764 provides review guidance for verifying that certain deterministic aspects of the change have been appropriately considered by the licensee. The criteria of this section include confirming that the licensee has provided adequate assurance that the change meets current regulations, except where specific exemptions are requested under 10 CFR 50.12. Additionally, criteria are also included for confirming that the licensee has provided adequate assurance that the change does not compromise defense-in-depth as it relates to the preservation of defenses against human errors.

3.1.2.3 Preservation of Sufficient Defense Against Human Errors

Section 4.1 of NUREG-1764 provides review guidance for verifying that certain deterministic aspects of the change have been appropriately considered by the licensee. This includes a criterion for providing adequate assurance that the change does not compromise defense-in-depth as it relates to the preservation of defenses against human errors.

The NRC staff reviewed the licensees description of defenses against human errors within the context of defense-in-depth. In section 3.5 of the LAR, the licensee stated that the proposed update to the Fire PRA does not impact the ability to rapidly detect, control, and promptly extinguish fires that do occur. The only aspect of the defense-in-depth approach that is altered is the timing associated with actions to stop RCPs to mitigate seal damage due to a total loss of RCP seal cooling. In the event of a complete loss of RCP seal cooling due to a fire, the previously identified recovery actions for defense-in-depth were determined to be sufficient and would be retained. No new recovery actions for defense-in-depth or for risk were determined to be necessary. The proposed update to the Fire PRA does not change the physical condition of the plant or its operations.

Based on the information in the LAR, as supplemented, the NRC staff finds that the preservation of defenses against human errors aspect of defense-in-depth conforms to the applicable criterion of NUREG-1764 and is acceptable to the extent needed to support this application because the manner in which defenses against human errors will be preserved provides adequate assurance that the change will not compromise defense-in-depth in this regard.

3.1.2.4 Analysis

Section 4.2, Analysis, of NUREG-1764 provides review guidance for verifying that the licensee has analyzed the changes to human actions and identified the HFE inputs for any modifications to procedures and training that may be necessary.

Functional and Task Analysis

The NRC staff reviewed the licensees descripti on of how personnel will know when the human action is necessary, as well as that it is performed correctly, including any changes to training or procedures. By letter dated August 25, 2023, the licensee confirmed that:

The action to trip the RCPs, which is only credited to occur from the control room, is currently proceduralized and frequently trained. Revisions to procedures for the RCP seal replacement are limited to those associated with changes in nomenclature or equipment IDs associated with key parameters that are monitored to assess the need for an RCP trip.

The licensee will also use procedures and training to help support the operators as described in section 3.6 below.

By letter dated August 25, 2023, the licensee further stated that:

The change in the seal configuration does not alter the method, the conditions and the priority of tripping the RCPs. The existing development for the human failure event modeling for failure to trip the RCPs includes and

addresses the factors specified in NUREG-1764, such as staffing, tasks, human-system interface, availability, accessibility, and training; and has been Peer Reviewed and found to meet, at Capability Category II, the associated requirements of the ASME/ANS PRA Standard. The primary impact of the RCP seal replacement on the human reliability analysis (HRA) is associated with the reduction in the system time window (Tsw). All other inputs to the HRA remain consistent with the current modeling. Therefore, the NUREG-1764 parameters are not directly impacted by the modification except for the system time window duration.

Additionally, the licensee stated that:

Since the total response time modeled (consisting of delay, cognitive and execution times) for this action is significantly shorter than the reduced system time window, the impact of the modification on the HEP is negligible.

The total delay, cognitive and execution time for this action, based on input from operator interviews, including a 5-minute additional delay for the value used in the Fire PRA, is less than 12 minutes. The system time window for the new seal configuration is 16 minutes.

Based on the information in the LAR, as supplemented, and the above discussion, the NRC staff finds that the task anal ysis conforms to the applicable criteria of NUREG-1764 and is acceptable to the extent needed to support this application because it describes how personnel will know when the human action is necessary and performed correctly, it describes the analysis of what the personnel must do and how human tasks are being changed, and it describes the analysis of potential errors and their consequences.

Staffing

The NRC staff reviewed the licensees descripti on of the effects of the changes in human actions upon the number, qualifications, and current staffing levels of operations personnel. By letter dated August 25, 2023, the licensee confirmed that there will be no changes to staffing as a result of the proposed modification.

Based on the information in the LAR, as supplemented, and the above discussion, the NRC staff finds that the licensees discussion of staffing changes conforms to the applicable criterion of NUREG-1764 and is acceptable to the extent needed to support this application because it describes the effects of changes in human actions upon the number and qualifications of operations personnel.

3.1.2.5 Procedures and Training

Section 4.3, Design of Human System-interfaces, Procedures, and Training, of NUREG-1764 provides review guidance for verifying that the licensee has supported the human actions by appropriate modifications to procedures and training.

Procedures

The NRC staff reviewed the licensees description of modifications to plant procedures as they relate to changes in operator task requirements. By letter dated August 25, 2023, the licensee confirmed that:

Revisions to procedures for the RCP seal replacement are limited to those associated with changes in nomenclature or equipment IDs associated with key parameters that are monitored to assess the need for an RCP trip. The update of plant procedures to reflect the changes associated with this modification are controlled by [Turkey Point] procedures which require that recommended procedure changes be identified in the Engineering Change package. Operations personnel review these recommended changes and initiate procedure changes for incorporation in conjunction with implementation of the plant modification. Operations training personnel review the design modification including the associated procedure changes and revise the appropriate operator lesson plans.

Based on the information in the LAR, as supplemented, and the above discussion, the NRC staff finds that the modifications to plant procedures conform to the applicable criterion of NUREG-1764 and are acceptable to the extent needed to support this application because they support the human actions by appropriate modifications to procedures.

Training

The NRC staff reviewed the licensees description of modifications to operator training as they relate to changes in operator task requirements. By letter dated August 25, 2023, the licensee confirmed that operations training personnel review the design modification including the associated procedure changes and revise the appropriate operator lesson plans.

Based on the information in the LAR, as supplemented, and the above discussion, the NRC staff finds that the training modifications conform to the applicable criterion of NUREG-1764 and are acceptable to the extent needed to support this application because they support the human actions by appropriate modifications to training.

3.1.2.6 Human Action Verification

Section 4.4, Human Action Verification, of NUREG-1764 provides review guidance for verifying that the licensee has demonstrated that the human actions can be successfully accomplished with the modified procedures and training.

By letter dated August 25, 2023, the licensee stated that [t]he validation time for the operator response to trip the RCPs following a loss of seal cooling is 6.17 minutes. This is much less than the 16 minutes assumed in the Fire PRA evaluation.

In reviewing the licensees validation results, the NRC staff supplemented the review criteria of NUREG-1764 with that of NUREG-1852. NUREG-1852 provides guidance for assessing the adequacy of time margins associated with the completion of manual operator actions.

Specifically, NUREG-1852, Appendix B, Summary of Expert Opinion Elicitation to Determine Time Margins for Operator Manual Actions in Response to Fire, section B.4, Identification of Time Margin and Conclusion, identifies a guideline that available time being at least

100 percent greater than the time obtained in a demonstration (i.e., a margin factor of at least 2) is supportive of a manual action being reliable. The NRC staff applied this methodology (i.e.,

Margin Factor = Allowed Time / Validated Time) to the licensees validation results and noted that the reduced allowed time of 16 minutes still yields a Margin Factor of 2.5 (16 minutes / 6.17 minutes = ~2.5).

Based on the information in the LAR, as supplemented, and the above discussion, the NRC staff finds that the walkthrough activities co nform to the applicable criterion of NUREG-1764 and NUREG-1852 and are acceptable to the extent needed to support this application because the validation time for the operator response to trip the RCPs is less, with sufficient margin, than the time of 16 minutes assumed in the Fire PRA evaluation and, therefore, the activities are described for determining that the human actions can be completed within the required time and are supported by procedures and training.

3.1.2.7 Human Factors Engineering Review Conclusion

The discussed general deterministic, analysis, procedures, training, and human action verification criteria collectively provide reasonable assurance that the licensee will comply with regulations and that the health and safety of the public will not be endangered. Therefore, the NRC staff finds that the facility control room will continue to conform to the requirements of Proposed GDC 11 with the facility design changes proposed in the LAR in place.

3.2 Probabilistic Risk Assessment Review

The NRC staff performed the following evaluations in accordance with RG 1.205 and RG 1.174 with respect to the proposed changes to the Turkey Point Fire PRA model to include the Framatome PSDS configuration and the refi nements provided in NUREG-2230, NUREG-2233, and NUREG-2178, Volume 2.

3.2.1 Evaluation of the Probabilistic Risk Assessment Acceptability The licensee stated that the Fire PRA (and the Internal Events PRA upon which it is based) has undergone formal industry peer reviews conducted by a diverse group of PRA practitioners from other plants and industry. The NRC staff also received confirmation from the licensee that there have been no upgrades made to the Turkey Point Internal Events or Fire PRA models since the last peer review. The NRC staff reviewed the Turkey Point PRA models and noted that these models have undergone several revisions to incorporate improvements and to maintain consistency with the as-built, as-operated plant.

The licensee adjusted the RCP seal PRA model to assess the risk impact of the use of the PSDS. Three PSDS seal failure modes, which are failure to actuate, failure to remain sealed, and spurious actuation, have been incorporated. Using the failure modes and effects analysis, the licensee developed fault trees that were used to quantify the probability of failure for each of the three failure modes. The licensee estimated the total seal failure probability as the sum of the individual probabilities of these three failure modes. The NRC staff noted that section 3.5 of to the LAR presents failure to actuate, failure to remain sealed, and spurious actuation of the PSDS as time-dependent failure rate frequencies. However, the development and evaluation of the failure modes was present ed in Attachment 3 to the LAR as per demand failures. Therefore, the NRC staff requested that the licensee provide additional information explaining the rationale and justification for the inconsistency. In its response by letter dated May 31, 2023, the licensee clarified that the seal failure configurations and the associated

failure modes are demand failure contributions to the seal failure value used in the PRA model.

The licensee confirmed that the values specified are per demand and consistent with the derived numbers in Attachments 3, 5, and 7 of the LAR. The NRC staff reviewed the fault trees associated with these PSDS failure modes to ensure that the methodology and assumptions are acceptable for this LAR for this licensee. The NRC staff also reviewed the associated human error probability of new operator actions added to the appropriate PRA models. The NRC staff did not perform an independent evaluation of the acceptability for generic use of the PSDS PRA models for other applications and licensees.

The NRC staff determined that the licensee used existing logics for the assessment and modeling of specific PSDS failure modes and properly applied the refinements in the above guidance as intended. Therefore, the NRC staff c oncludes that the PRA models used for this LAR are acceptable.

3.2.2 PRA Results

RG 1.205 states that an evaluation should be performed to address the acceptability of the change in terms of the plant change evaluation criteria and compliance with the fire protection requirements of 10 CFR 50.48(c). The plant change evaluation process should also include an integrated assessment of the acceptability of the change in risk, defense-in-depth, and safety margins, regardless of the methods or approaches used to evaluate the change. RG 1.174 provides acceptance guidance applicable to NFPA 805 plant change evaluations.

The licensee stated that the fire risk evaluation process was performed in accordance with RG 1.205, with acceptance criteria provided by RG 1.174. However, the NRC staff noted that there was no estimate for seismic and other external hazards. Therefore, the NRC staff asked the licensee to provide seismic and other external hazards contribution to CDF/LERF as appropriate. In its response by letter dated July 27, 2023, the licensee provided an updated PRA result with recent seismic refinements, which resulted in a delta CDF of 4.30 x 10-06/yr for Unit No. 3 and a delta CDF of 4.61 x 10-06/yr for Unit No. 4. The delta LERFs were calculated to be 3.00 x 10-08/yr and 4.00 x 10-08/yr for Unit Nos. 3 and 4, respectively. The corresponding total plant CDFs are 7.05 x 10-05/yr for Unit No. 3 and 7.58 x 10-05/yr for Unit No. 4, and the total plant LERFs are 1.59 x 10-06/yr for Unit No. 3 and 1.61 x 10-06/yr for Unit No. 4. These results fall within Region II on both Figure 4, Acceptance guidelines for core damage frequency, and Figure 5, Acceptance guidelines for large early release frequency, of RG 1.174, which are considered to be small changes per the acceptance guidelines. Therefore, the NRC staff determined that the change is permissible and is consistent with the intent of the Commissions Safety Goal Policy Statement (51 FR 28055, August 4, 1986, as corrected and republished at 51 FR 30028, August 21, 1986).

3.2.3 Evaluation of Defense-in-Depth and Safety Margin

RG 1.205 specifies that proposed changes to the plant must also be consistent with the defense-in-depth philosophy and maintain sufficient safety margins. Fire protection defense-in-depth is achieved when an adequate balance of each of the following elements is provided:

(1) Preventing fires from starting.

(2) Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage; and

(3) Providing an adequate level of fire protection for structures, systems, and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

The NRC staff determined that installing the Framatome PSDS and updating the Fire PRA as proposed in the LAR would not impact the ability to prevent fires from starting, or impact the ability to rapidly detect, control, and promptly extinguish fires that do occur. The only aspect of defense-in-depth affected by the proposed change is the timing associated with the action to stop the RCPs to mitigate seal damage due to a total loss of RCP seal cooling. The NRC staff noted that with the new Framatome PSDS design, plant operators would have to trip the RCPs within 16 minutes during an LOSC event to prevent RCP seal damage. This is notably less available time than the 20 minutes afforded by the currently installed Flowserve RCP seal design. Therefore, the NRC staff requested that the licensee provide additional discussion of the factors that led to the reduced time available to perform the operator manual actions, and the potential impact on the feasibility and reliability of performing the required actions in a timely manner.

In its response by letter dated May 31, 2023, the licensee stated that the operator action to trip the RCPs is an action in the main control room and is, therefore, not a recovery action but a required post-fire operator action taken in the control room. In addition, the licensee stated that the associated human error probability of this new operator action has been added to the appropriate PRA models and the action has been incorporated into appropriate operating procedures. Since the action to trip the RCPs is performed in the main control room and has been incorporated into plant procedures, the NRC staff concludes that this action can reasonably be expected to be achievable within 16 minutes.

RG 1.174 states that sufficient safety margins are maintained when:

(1) The codes and standards or their alternatives approved for use by the NRC are met; and

(2) Safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met or proposed revisions provide sufficient margin to account for uncertainty in the analysis and data.

The licensee stated in the LAR that the fire risk evaluations and change evaluations utilized for this application are consistent with the methodology previously used for the NFPA 805 performance-based approach and in accordance with RG 1.205. The Fire PRA model update to include the refinements provided in NUREG-2230, NUREG-2233, and NUREG-2178, Volume 2, are also per NRC guidelines. The licensee also stat ed that a review of the Turkey Point UFSAR, Chapter 14 safety analysis has determined that two Loss of Reactor Coolant Flow scenarios rely on the ability of the RCPs to coastdown which is followed by natural circulation of the reactor coolant to remove residual and decay heat. Since RCP coastdown is independent of RCP seal operation, the safety analyses acceptance criteria are not affected by the proposed modification. Based on the above, the NRC sta ff concludes that the licensee has used acceptable approaches in updating PRA models in support of this LAR and that the changes do not affect the safety analyses acceptable criteria in Turkey Points licensing basis.

In consideration of the above discussions, the NRC staff concludes that defense-in-depth and adequate safety margins continue to be maintained for the Turkey Point FPP per RG 1.205 and RG 1.174.

3.2.4 Evaluation of Performance Monitoring

RG 1.174 provides guidance for performance m onitoring to ensure that no unexpected adverse safety degradation occurs because of the proposed change to the licensing basis. The licensee stated that the Framatome PSDSs have a 12-year operating life with a 6-year inspection period.

This mid-service PSDS inspection ensures that the PSDS has not inadvertently actuated and is in normal condition. During seal replacement, the PSDS may be inspected for scoring, scratches, raised metal, pits, or chips. Any unusual wear or changes in operational characteristics will be addressed in the plants corrective action program. During operation with the new Framatome RCP hydrostatic seals wi th PSDS, performance parameters including the seal No. 1 leak-off flow and temperature would be monitored.

The monitoring of RCP performance parameters during operation along with the initial and periodic inspection results will inform future Framatome PSDS inspection intervals. Therefore, although the use of the Framatome PSDS presented a small risk increase, the NRC staff determined that the above performance monitoring strategy will limit the potential risk and provide assurance that degrading performance can be detected and corrected in a timely manner.

Based on the above discussion, the NRC staff concludes that the performance monitoring strategy to be implemented by the licensee is adequate for the small risk increase in accordance with RG 1.174.

3.2.5 PRA Review Conclusion

Based on the above, the NRC staff concludes that the proposed change is acceptable because the PRA analyses supporting the LAR meet the requirements of RG 1.200, the increase in CDF and LERF as the result of the proposed change is acceptable per RG 1.174, defense-in-depth and safety margins are maintained, and the performance of the seals will be monitored.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Florida State official was notified of the proposed issuance of the amendments on Nove mber 11, 2023. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the 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 any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on December 27, 2022 (87 FR 79356) 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 impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 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.

Principal Contributors: T. Dinh, NRR C. Moulton, NRR C. Ng, NRR G. Bedi, NRR T. Scarbrough, NRR M. Keefe-Forsyth, NRR S. Sun, NRR M. Mahoney, NRR

Date: January 22, 2024

ML23320A306 NRR-058 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DRA/APLB NAME MMahoney RButler (ABaxter for) JWhitman DATE 12/04/2023 11/30/2023 09/27/2023 OFFICE NRR/DEX/EMIB NRR/DRO/IRAB NRR/DSS/SNSB NAME SBailey PFinney PSahd DATE 12/12/2023 12/06/2023 12/05/2023 OFFICE OGC - NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME RSiegman DWrona MMahoney DATE 01/08/2023 01/22/2023 01/22/2023