ML23235A296

Issuance of Amendment Nos. 247 and 244, Regarding License Amendment Request to Revise Technical Specification 3.6.5, Containment Air Temperature, (EPID L-2023-LLA-0116) (Emergency Circumstances)

ML23235A296
Person / Time
Site:	Farley
Issue date:	08/24/2023
From:	Dawnmathews Kalathiveettil NRC/NRR/DORL/LPL2-1
To:	Coleman J Southern Nuclear Operating Co
References
EPID L-2023-LLA-0116
Download: ML23235A296 (27)

Text

August 24, 2023 Ms. Jamie Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENT NOS. 247 AND 244, REGARDING LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE, (EPID L-2023-LLA-0116) (EMERGENCY CIRCUMSTANCES)

Dear Ms. Coleman:

The Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 247 to Renewed Facility Operating License No. NPF-2 and Amendment No. 244 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated August 22, 2023, as supplemented by letters dated August 23, 2023.

The amendments revise the operating license and approve a one-time NOTE to Appendix A TS Limiting Condition for Operation (LCO) 3.6.5, Containment Air Temperature, to revise the limit on containment average air temperature from 120°F to 122°F (Fahrenheit) effective until 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> Central Time on September 9, 2023.

These amendments are issued under emergency circumstances as described in the provisions of paragraph 50.91(a)(5) of Title 10 of the Code of Federal Regulations due to the time critical nature of the amendment.

In this instance, an emergency situation exists due to the unforeseen emergent conditions of the Farley site ambient temperature being persistently elevated including the upcoming period of projected temperatures to exceed approximately 100°F for multiple days. This ambient condition is projected to result in both Farley Units 1 and 2 containment average air temperatures exceeding 120°F. Additionally, the Southeastern Reliability Coordinator (SeRC) has elevated a Conservative System Operations (CSO) Watch to a CSO Warning, applicable to both Farley units.

J. Coleman A copy of the related safety evaluation is also enclosed. The safety evaluation describes the emergency circumstances under which the amendment was issued and the final no significant hazards determination. A Notice of Issuance addressing the final no significant hazards determination and opportunity for a hearing associated with the emergency circumstances will be included in the Commissions 28-Day Monthly Federal Register notice.

If you have questions, you can contact me at 301-415-5905 or at Dawnmathews.Kalathiveettil@nrc.gov.

Sincerely,

/RA/

Dawnmathews T. Kalathiveettil, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

1. Amendment No. 247 to NPF-2

2. Amendment No. 244 to NPF-8

3. Safety Evaluation cc: Listserv

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 247 Renewed License No. NPF-2

1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by Southern Nuclear Operating Company, Inc.

(Southern Nuclear), dated August 22, 2023, as supplemented by letters dated August 23, 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 license 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, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-2 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 247, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented immediately.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Digitally signed by Michael T. Markley Date: 2023.08.24 18:35:21 -04'00' Markley Michael T. Markley, Chief Plant Licensing Branch II-1 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 24, 2023

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 244 Renewed License No. NPF-8

1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by Southern Nuclear Operating Company, Inc.

(Southern Nuclear), dated August 22, 2023, as supplemented by letters dated August 23, 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 license 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, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-8 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 244, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented immediately.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Digitally signed by Michael T. Markley Date: 2023.08.24 18:36:13 -04'00' Markley Michael T. Markley, Chief Plant Licensing Branch II-1 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 24, 2023

ATTACHMENT TO JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT NO. 247 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND LICENSE AMENDMENT NO. 244 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License License NPF-2, page 4 NPF-2, page 4 NPF-8, page 3 NPF-8, page 3 TSs TSs 3.6.5-1 3.6.5-1

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 247, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the Issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.

a. Southern Nuclear shall not operate the reactor in Operational Modes 1 and 2 with less than three reactor coolant pumps in operation.

b. Deleted per Amendment 13

c. Deleted per Amendment 2

d. Deleted per Amendment 2

e. Deleted per Amendment 152 Deleted per Amendment 2

f. Deleted per Amendment 158

g. Southern Nuclear shall maintain a secondary water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include:

1) Identification of a sampling schedule for the critical parameters and control points for these parameters;

2) Identification of the procedures used to quantify parameters that are critical to control points;

3) Identification of process sampling points;

4) A procedure for the recording and management of data;

5) Procedures defining corrective actions for off control point chemistry conditions; and Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 247

(2) Alabama Power Company, pursuant to Section 103 of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess but not operate the facility at the designated location in Houston County, Alabama in accordance with the procedures and limitations set forth in this renewed license.

(3) Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproducts, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporate below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 2775 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 244, are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3) Delete per Amendment 144 (4) Delete Per Amendment 149 (5) Delete per Amend 144 Farley - Unit 2 Renewed License No. NPF-8 Amendment No. 244

Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be 120°F.

---

NOTE----------------------------------------------

Containment average air temperature shall be 122°F until 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on September 9, 2023.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A.1 Restore containment 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> temperature not within average air temperature limit. to within limit.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within In accordance with limit. the Surveillance Frequency Control Program Farley Units 1 and 2 3.6.5-1 Amendment No. 247 (Unit 1)

Amendment No. 244 (Unit 2)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 247 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 244 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364

1.0 INTRODUCTION

By application dated August 22, 2023, (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML23234A151), as supplemented by letters dated August 23, 2023 (ML23235A176 and ML23235A288), Southern Nuclear Operating Company, Inc. (SNC, the licensee) requested changes to the Technical Specifications (TSs) for Renewed Facility Operating License Nos. NPF-2 and NPF-8 for Joseph M. Farley Nuclear Plant (Farley),

Units 1 and 2, respectively. The supplement letters were provided in response to the NRC request for additional information (RAI) on August 22, 2023 (ML23236A002).

The proposed changes revise the operating license and approve a one-time NOTE to Appendix A Limiting Condition for Operation (LCO) 3.6.5, Containment Air Temperature, to revise the limit on containment average air temperature from 120°F to 122°F effective until 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> Central Time on September 9, 2023.

The licensee requested U.S. Nuclear Regulatory Commission (NRC) approval of the proposed amendment in accordance with Section 50.91(a)(5) of Title 10 of the Code of Federal Regulations (10 CFR) regarding emergency situations, as discussed in Section 4.0 of this safety evaluation.

1.1 System Description In Section 2.1 of the Enclosure to its submittal dated August 22, 2023, the licensee stated:

The containment is a prestressed, reinforced concrete cylindrical structure with a shallow domed roof and a reinforced concrete foundation slab. A 1/4-in.-thick Enclosure 3

welded steel liner is attached to the inside face of the concrete. The floor liner is installed on top of the foundation slab and is then covered with concrete. The containment completely encloses the reactor, the reactor coolant systems, the steam generators, and portions of the auxiliary and engineered safeguards systems. It ensures that an acceptable upper limit for leakage of radioactive materials to the environment will not be exceeded even if gross failure of the reactor coolant system occurs. The structure is designed to contain radioactive material that may be released from the reactor core following a Design Basis Accident (DBA). Additionally, this structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.

As described in FSAR [Final Safety Analysis Report] subsection 6.2.2, three systems are provided to reduce containment atmosphere temperature and pressure and/or to remove heat from the containment under post-accident conditions. These are the low-head safety injection/residual heat removal system, the containment spray system, and the containment cooling system.

The two redundant trains of the containment spray system have been designed to provide sufficient heat removal capacity to prevent exceeding containment design pressure for all piping breaks. The containment cooling system has been designed to remove heat which will be released to the containment atmosphere during any Main Steam Line Break (MSLB) or Loss-of-Coolant-Accident (LOCA) up to and including the double-ended rupture of the largest system pipe. This is accomplished by one of four containment air coolers.

As described in FSAR subsection 6.2.1.3.3, Containment Pressure Transient Analysis, and shown in Table 6.2-3, Initial Conditions for Pressure Analysis, and Table 6.2-19, Containment Results for the Design Basis LOCA, the analyses for containment pressure assumed an initial containment temperature of 127°F.

1.2 Proposed Change Current TS LCO TS 3.6.5 requires, Containment average air temperature shall be 120°F. If the LCO 3.6.5 is not met, under Condition A, the plant has 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to restore the temperature within limits. If the Condition A is not met, Condition B, Required Action B.1 requires the plant must be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Required Action B.2 requires the plant to be in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Revised TS LCO 3.6.5 would add a one-time NOTE that would state:

Containment average temperature shall be 122°F until 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on September 9, 2023.

The NRC staff notes that Farley Nuclear Plant is in a Central Time zone.

2.0 REGULATORY EVALUATION

The applicable regulatory requirements and guidance is provided in the following subsections:

2.1 Applicable Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) Part 50.90, Application for amendment of license, construction permit, or early site permit, requires that whenever a holder of a license wishes to amend the license, including TSs in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate.

The regulations in 10 CFR 50.91(a)(5) states, in part, that where the Commission finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plants licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment.

Under 10 CFR 50.36(c)(2), TSs must contain Limiting Conditions for Operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee must shut down the reactor or follow any remedial action permitted by the TSs until the LCO can be met.

Typically, the TSs require restoration of equipment in a timeframe commensurate with its safety significance, along with other engineering considerations. Under 10 CFR 50.36(b), TSs must be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.

The regulations in 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, requires, in part, licensees to establish a program for qualifying the electric equipment important to safety. The electric equipment under the scope of this section includes safety-related equipment, non-safety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified by the safety-related equipment, and certain post-accident monitoring equipment. 10 CFR 50.49(e)(1) requires the time-dependent temperature and pressure at the location of the electric equipment important to safety must be established for the most severe design-basis accident during or following which this equipment is required to remain functional.

In determining whether the proposed TS remedial actions should be granted, the Commission will apply the reasonable assurance standards of 10 CFR 50.40(a) and 50.57(a)(3). The regulation at 10 CFR 50.40(a) states that in determining whether to grant the licensing request, the Commission will be guided by, among other things, consideration about whether the processes to be performed, the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals, in regard to any of the foregoing collectively provide reasonable assurance that the applicant will comply with the regulations in this chapter, including the regulations in [Part] 20 of this chapter, and that the health and safety of the public will not be endangered. The regulation at 10 CFR 50.57(a)(3) states that the Commission may issue an operating license when it has, in part, reasonable assurance that the activities authorized by the operating license may be conducted without endangering the health and safety of the public.

The regulations in 10 CFR 50, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components (SSCs) important to safety. The applicable GDCs for this submittal includes:

Criterion 4 [GDC 4], Environmental and dynamic effects design bases, requires that structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

Criterion 16 [GDC 16], Containment design, requires that the reactor containment and associated systems shall be provided to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

Criterion 38 [GDC 38], Containment heat removal, specifies that a system to remove heat from the reactor containment shall be provided that rapidly reduces, consistent with the functioning of other associated systems, the containment pressure and temperature following any LOCA and maintains them at acceptable low levels.

Criterion 50 [GDC 50] Containment design basis, requires that the reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartment can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from a design basis LOCA.

2.2 Licensing Basis Document The Farley FSAR Section 3.11, Environmental Design of Mechanical and Electrical equipment, states:

The original specifications for safety-related electrical equipment which is subject to a post DBA harsh environment and required to function during and subsequent to a DBA required qualification to [Institute of Electrical and Electronics Engineers] IEEE 323-1971, [IEEE Trial-Use Standard:

General Guide for Qualifying Class I Electric Equipment for Nuclear Power Generating Stations.] Subsequently, the Farley Nuclear Plant Environmental Qualification (EQ) Program was implemented to comply with the requirements of NRC Inspection and Enforcement Bulletin (IEB)79-01B [Environmental Qualification of Class IE Equipment,]

NUREG-0588, Revision 1, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, [ML031480402] and 10 CFR 50.49. Based on the dates of the Farley plant operating licenses, Unit 1 was required to comply with the requirements of IEB 79-01B, which

provides the NRC Division of Operating Reactors (DOR) Guidelines, and Unit 2 was required to comply with the requirements of NUREG-0588, Category II. The requirements set forth under these programs supplement the requirements of IEEE 323-1971 [IEEE Trial-Use Standard: General Guide for Qualifying Class I Electric Equipment for Nuclear Power Generating Stations.] After implementation of these programs, 10 CFR 50.49 was issued and mandated environmental qualification requirements for safety related electrical equipment. Regulatory Guide 1.89, Revision 1,

[Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, ML003140271] followed and established IEEE 323-1974, [IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations] as an acceptable standard to comply with the requirements of 10 CFR 50.49. The provisions of 10 CFR 50.49 waive the need to requalify components previously qualified under the DOR Guidelines or NUREG-0588 unless the components are replaced.

The replacement components must comply with the provisions of 10 CFR 50.49 unless there are sound reasons to the contrary. These reasons, when required, will be documented. Accordingly, the EQ program implements the requirements of 10 CFR 50.49 as documented in the EQ master lists and the associated EQ packages. The EQ packages document which version of the IEEE-323 standard was used for the qualification.

2.3 Applicable Regulatory Guidance Regulatory Guide (RG) 1.89, Revision 1, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants (ML003740271), describes a method acceptable to the NRC staff for complying with 10 CFR 50.49 with regard to qualification of electric equipment important to safety for service in nuclear power plants to ensure that the equipment can perform its safety function during and after a design-basis accident.

Inspection and Enforcement Bulletin (IEB)79-01B, Environmental Qualification [EQ] of Class 1E Equipment (NRC microfiche 7910250528), required the licensee to perform a detailed review of the environmental qualification of Class 1E electrical equipment to ensure that the equipment will function under (i.e., during and following) postulated accident conditions.

NUREG-0588, Revision 1, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment (ML031480402), provides interim staff positions on selected areas of environmental qualification of electrical equipment including guidance on (1) how to establish environmental service conditions, (2) how to select methods which are considered appropriate for qualifying the equipment in different areas of the plant, and (3) other areas such as margin, aging, and documentation.

Section 3.11 of NUREG-800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, provides guidance on EQ of mechanical and electrical equipment for complying with 10 CFR 50.49.

RG 1.155, Station Blackout, describes a means acceptable to the NRC staff for meeting the requirements of 10 CFR 50.63, Loss of all alternating current power. NUMARC-8700, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light

Water Reactors (ML12137A732), also provides guidance acceptable to the staff for meeting these requirements. NUMARC-8700, Section 2.7.1, identifies the assumptions associated with loss of ventilation and indicates Equipment Operability Inside Containment Temperatures resulting from the loss of ventilation are enveloped by the loss of coolant accident (LOCA) and high energy line break environmental profiles.

NUREG-0800, Chapter 6, Engineered Safety Features.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the emergency amendment request to determine whether the proposed changes are consistent with the regulatory requirements, the plant-specific design and the licensing basis. The NRC staff evaluations are detailed in the following sub-sections:

3.1 Containment DBA Analysis The containment average air temperature limit is an initial condition for the DBA analysis to ensure that the containment spray and containment cooling system can maintain the total amount of energy in the containment within the specified limits and are able to remove heat from the containment during post-accident conditions. It is an important consideration in establishing the containment environmental qualification operating envelope for both pressure and temperature.

The LCO 3.6.5 for containment average air temperature limit ensures that the initial conditions assumed for the containment response for a MSLB or LOCA event are not violated during the plant operations. During the post-accident conditions, the total amount of energy removed from the containment by the containment spray and the containment cooling system is dependent upon the energy released into the containment due to the event. Exceeding the containment design limits for pressure and temperature may result in leakage rates greater than that assumed for the accident analysis.

Mass and Energy Release and Containment Response The containment analyses performed for maximum temperature and pressure for the MSLB and LOCA assume an initial bulk containment temperature of 127 °F. In its supplement to the emergency LAR (ML23235A176), the licensee stated the total channel uncertainty for the containment average air temperature is 2.5 percent. Thus, the proposed TS LCO limit of 122 °F is bounded by the safety analysis limit of 127 °F. The licensee used the square root of the sum of the squares of independent components method to calculate the instrument uncertainty. The licensee stated in the supplement that the methodology used to calculate the instrumentation uncertainty is consistent with the previously NRC accepted methods. The NRC staff independently reviewed the method for calculating the instrument uncertainty and finds it followed NRC-approved methods. Based on the above, the NRC staff concludes there is reasonable assurance that the increase in containment average air temperature limit, along with uncertainty applied, remains bounded by the 127°F analysis of record and is, therefore, acceptable.

The LOCA analysis for Farley assumed an accumulator liquid temperature of 120°F and the Refueling Water Storage Tank (RWST) initial temperature of 110°F. As the containment air temperature increases, the accumulator temperature will also increase. To accommodate the increase of accumulator temperature to 122°F, the licensee performed calculations assuming an

RWST initial temperature decrease from 110°F to 100°F. The current operation data at Farley shows the RWST to be below 95°F. In its supplement (ML23235A176), the licensee stated that administrative tracking is in place to require control room operators to verify the RWST temperatures remains less than 100°F. Based on the above and administrative controls, the NRC staff finds the decrease in RWST initial temperature used for the LOCA analysis to be acceptable.

The licensee performed calculations to show that the increase in energy due to a 2°F increase in accumulator liquid temperature is sufficiently offset by a decrease in integrated break energy from the 10°F decrease in the RWST initial temperatures and would result in a net decrease in the energy into the containment. Based on its independent evaluation of the licensees analysis, the NRC staff finds the impact on containment pressure from a potential increase in accumulator temperature for the LOCA analysis to be acceptable. Based on its evaluation of the licensees analysis and review of the containment design limits for pressure and temperature in Section 6.2 of the UFSAR (ML21314A283), the NRC staff concludes there is reasonable assurance that anticipated increase in the containment vapor temperature, sump temperature, or containment peak pressure following a MSLB or LOCA remain within the analysis of record and are, therefore, acceptable.

Based on the above, the NRC staff concludes that the analytical methods for the channel uncertainty calculation and for the use of lower initial RWST temperature is sufficient to provide reasonable assurance that the proposed changes would remain within the analysis of record for the limited period of the one-time request that is effective until 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on September 9, 2023.

Small Break LOCA Evaluation A small break LOCA (SBLOCA) for Farley is defined as a breach in the reactor coolant pressure boundary with a total cross-sectional area less than 1.0 ft2. As previously discussed, the increase in the containment air temperature leads to an increase in the accumulator temperature. The licensee calculated that a 2°F increase in accumulator temperature leads to an enthalpy increase of ~2 Btu/lbm, which corresponds to a small reduction in total energy removal capability of the accumulator fluid. The licensee concluded that the proposed temperature increase will have a negligible impact on SBLOCA analysis of record due to (1) a relatively small reduction in energy removal capability of the accumulator fluid, (2) the low amount of stored energy in the core and vessel internals for the transient, and (3) the accumulator injection characteristics remaining unaffected.

The NRC staff independently evaluated the licensees analysis and concludes there is reasonable assurance that there will be negligible impact on peak cladding temperature (PCT) and maximum local oxidation on cladding during a SBLOCA due to the proposed change and would, therefore, remain bounded by the analysis of record and is, therefore, acceptable.

Net Positive Suction Head (NPSH) Evaluation In its submittal dated August 22, 2023, the licensee stated that there is 14 feet or greater head available for up to 180°F containment sump temperature. As the sump temperature increases above 140°F, the strainer head losses decrease. The licensee states that due to the competing effect of decreasing strainer head loss and the sump vapor pressure, the pump Net Positive Suction Head (NPSH) margin would be expected to increase or stay the same as the sump temperature increases above 212°F.

The NRC staff independently evaluated the licensees analysis and concludes there is reasonable assurance that the NPSH margin will remain unchallenged by the increase in containment air temperature from 120°F to 122°F and is, therefore, acceptable.

3.2 Service Water (SW) and Ultimate Heat Sink (UHS) Evaluation In its submittal dated August 22, 2023, the licensee states that the Service Water (SW) discharge from both Farley units goes to the ultimate heat sink (UHS) pond absorbing decay heat from both units and considers various combinations for shutdown and accident conditions for the 30-day period. The licensee stated that small increase in containment temperature at the start of the event represents an insignificant effect compared to the magnitude of decay heat from both units. As a basis for the emergency LAR, the licensee stated that August average temperatures have been higher compared to historical averages and cited lower than historical rainfall area for the month of August. The higher SW pond temperatures due to rising surrounding temperatures could lead to further challenges to the licensee requested 122°F containment air temperature limit; however, in the supplement to the LAR (ML23235A176), the licensee provided recent trend information for the SW temperature values which remain below the TS Surveillance Requirement 3.7.9.2 limit of 95°F. Based on its independent evaluation, the NRC finds staff finds that containment air temperature will have negligible impact on SW pond temperatures and the UHS. Based on the recent trend for values provided for the SW temperatures, the NRC staff concludes the increased containment air temperature limit of 122°F would provide reasonable assurance that the SW and UHS would remain within the analysis of record.

3.3 Post LOCA Evaluation Subcriticality Assessment The increase in maximum accumulator temperature will change the accumulator mass slightly due to the density decrease. The licensee stated that the change in density due to increase in temperature from 120°F to 122°F is small enough to not have any measurable impact on the accumulator mass used in the subcriticality calculations.

Based on its independent evaluation, the NRC staff finds that the small change in accumulator temperature will maintain the accumulator mass within the analysis of record. The NRC staff notes that an increase in accumulator temperature will lead to an increase in the solubility of the boric acid and would therefore, have some favorable impact on the subcriticality analysis.

Sump Dilution and Hot Leg Switchover Assessment The licensee states that the accumulator mass calculation is based on a higher density value and is not dependent on the maximum accumulator temperature. Hence, the licensee states that there is no impact on post LOCA sump dilution calculation due to the small increase in accumulator temperature from 120°F to 122°F. Further, as stated in the previous section, the density decrease due to an increase in temperature from 120°F to 122°F is small enough to not have any measurable impact on the accumulator mass. The licensee also states that the hot leg switchover analysis uses the same accumulator mass value used in the sump dilution calculation and thus is not impacted by the small increase in accumulator temperature.

Based on the above, the NRC staff concludes that there is reasonable assurance that the proposed changes are bounded by the plant analysis of record and is, therefore, acceptable for the limited period of the one-time request.

Decay Heat Removal Assessment In its letter dated August 22, 2023, the licensee stated that any minor changes in core voiding and core boil-off rates resulting from the 2°F accumulator temperature increase are relatively short-term effects that do not persist into the long-term cooling phase of the emergency core cooling system (ECCS) performance evaluations.

Based on the above, the NRC staff concludes there is reasonable assurance that any changes in the core voiding or boil-off rates due to increase in accumulator temperature from 120°F to 122°F are short term effects and thus do not impact the long-term cooling phase of the Emergency Core Cooling System (ECCS) and is, therefore, acceptable.

3.4 Large Break LOCA PCT Evaluation The maximum accumulator temperature is a parameter used as input in large break LOCA (LBLOCA) analysis. If the accumulator temperature is higher during an LBLOCA, fuel temperatures will increase for a longer period. This will ultimately cause the PCT to be higher.

When the containment air temperature increases, the accumulator temperature also increases since the accumulator is located inside the containment. In its letter dated August 22, 2023, the licensee stated:

Accumulator temperature sensitivities from similar pressurized water reactor plant designs with similar fuel assembly design, power level, were performed and predicted cladding temperature response to determine an estimated effect for FNP. It is acknowledged that accumulator temperature sensitivities were executed prior to modeling of fuel thermal conductivity degradation (TCD) in best-estimate LOCA analyses and fuel performance codes. The sensitivities remain valid for the purpose of estimating the effect of the increase accumulator temperature range; however, the overall PCT for limiting transients when modeling fuel TCD are on the order of 150°F higher. As such, a conservative multiplier of 2 is applied to the estimate of effect to account for the use of pre-TCD transient results from representative plants for the Farley evaluation.

The estimated effect is a PCT increase of ~0.5°F per 1°F increase in accumulator temperature. As such, with the conservative multiplier, the 2°F increase in the maximum accumulator temperature is estimated to have a 2°F effect on the analysis PCT. The latest racked up PCT is 2034°F. With the additional 2°F, the racked up PCT is estimated to be 2036°F, which maintains margin to the regulatory acceptance criterion of 2200°F.

The NRC staff independently evaluated the licensees analysis and compared it to similar licensing actions and found that the LBOCA PCT impact due to a relatively small rise in accumulator temperature could exceed the 2°F, as stated by the licensee, but expect that it would be less than 100°F. One plant, for example, R.E. Ginna Nuclear Power Plant performed calculations to increase its containment average air temperature 5°F increase from 120°F to 125°F (ML14232A331). The current Farley LBLOCA rackup PCT for the licensee is 2034°F, so there is a 166°F margin to the 2200°F PCT limit in 10 CFR 50.46(b)(1). Based on the above, there would still be sufficient margin to the regulatory limit even if a conservative increase in LBLOCA PCT of 100°F is assumed. Additionally, the licensee provided calculations for the expected decrease in accumulator heat removal capability due to the 2°F increase in accumulator temperature and stated that it would be more than offset by the margin present in the RWST temperature and corresponding heat removal temperature, as discussed in the

Section 3.1 of this SE.

The NRC staff finds the 2°F increase in containment temperature (and thus accumulator temperature) to be acceptable because (1) there is enough margin in the current LBLOCA rackup PCT to accommodate a conservative estimate on the PCT impact; (2) the margin to the RWST temperature would be expected to offset the increase in accumulator temperature and (3) the LAR for increased containment temperature being temporary and applicable for the duration effective until 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on September 9, of 2023. Based on the above, the NRC staff finds there is reasonable assurance that the proposed changes are bounded by the plant analysis of record and is, therefore, acceptable for the limited period of the one-time request.

3.5 Generic Letter (GL) 96-06, Assurance of Equipment Operability and Containment Integrity during Design Basis Accident Conditions In its supplement to the LAR (ML23235A176), the licensee stated:

SNC response to GL 96-06 continues to be based on the current calculation of record and remains unaffected. The current Farley analyses determine that the water hammer pressure spikes at the containment air coolers are not significant enough to cause damage, demonstrating that Farley is not susceptible to the Generic Letter 96-06 concerns. A containment air temperature increase from 120°F to 122°F has a negligible impact on the current analysis. The waterhammer analysis identified that the region of the service water piping that is most susceptible to water hammer is the containment cooler return piping with lowest possible system backpressure. However, this analysis showed that no waterhammer will occur within this piping. Following a LOCA coincident with a LOSP [loss-of-offsite power], both the containment cooler fans and service water pumps are de-energized resulting in reduced air and service water flows through the coolers. Therefore, the heat transfer from the containment atmosphere will be less than the full capacity of the containment coolers. In the time interval of interest (25 seconds or less) following initiation of this event, the service water downstream of the containment coolers has been calculated to reach a maximum of 119°F. Even if the service water downstream of the containment coolers picks up an additional 2°F, the maximum temperature will still be less than the 164°F temperature required to form a vapor cavity.

Service water temperature, along with its associated vapor pressure, will rise rapidly following containment cooler fan restart. However, the increase in the service waters vapor pressure, caused by the service waters temperature rise, will not reach the increased pressure in the containment cooler discharge piping. Therefore, two phase flow conditions still will not occur.

Based on its independent evaluation of the licensee-provided justification, the NRC staff concludes there is reasonable assurance that Farley is not susceptible to the GL 96-06 (ML031110029) concerns.

3.6 Equipment Qualification of Electrical Equipment Important to Safety NRC staff reviewed the submittal to determine the impact of the proposed change on the electrical equipment within containment. In the emergency LAR, the licensee noted a temporary 2°F increase in the containment average air temperature is bounded by the existing equipment qualification analyses, due to conservatisms and margins in the existing test

programs and calculations and that the increase in containment average temperature limit will have no impact on the qualification status or qualified lives of existing equipment located in containment.

On August 22, 2023 (ML23236A002), NRC staff requested additional information (RAI) from the licensee to clarify (1) its use of the term equipment qualification and to identify the requirements the it implied will continue to be met and (2) clarify whether there are any non-environmentally qualified electric equipment important to safety within containment that is expected to perform a design function under normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

In its response to the staffs RAI, the licensee stated:

In accordance with FNP Environmental Qualification Program, the program scope includes electrical equipment that is important to safety. Equipment that is important to safety involves safety related and non-safety related electrical equipment whose failure can prevent satisfactory accomplishment of safety functions as described in 10 CFR 50.49 (b)(1) and (b)(2) and certain post-accident monitoring equipment as described in 10 CFR 50.49(b)(3).

Based on the above, the NRC staff finds that there are no changes in the environmental qualification of equipment in containment and that the licensees analysis of record remains bounding in accordance with 10 CFR 50.49.

Based on the above, the NRC staff finds that there remains sufficient margin in the environmental qualification program and the environmental qualification of electric equipment at Farley should not be adversely impacted by the proposed 2°F increase in containment temperature. Based on the above, the NRC staffs independent evaluation concluded that there is reasonable assurance that the proposed changes are bounded by the plant analysis of record for equipment environmental qualification and is, therefore, acceptable for the limited period of the one-time request.

3.7 Evaluation of Technical Specifications Change Regulations in 10 CFR 50.36(c)(2) provides requirements for TSs LCO. Specifically, 10 CFR 50.36(c)(2)(ii)(B), Criterion 2, requires that an LCO be established for, A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCO 3.6.5, Containment Air Temperature, establishes a limit on the average containment air temperature to assure that the assumptions for accident analyses associated with loss of coolant and steam line break accidents are valid. A containment air temperature higher than that allowed by the LCO could result in a plant response to a DBA that is more severe than that calculated in the analyses.

The licensee proposes to add a note to LCO 3.6.5 that allows the maximum average containment air temperature to temporarily be increased by 2 degrees Fahrenheit. As discussed in this SE, the proposed increase in average containment air temperature will not result in consequences more severe than those already evaluated in the plant safety analyses. With the addition of the proposed note, LCO 3.6.5 also continues to meet the requirements of 10 CFR 50.36(c)(2)(i) by establishing the requirements for the lowest functional capability of the containment and mitigating systems required for safe operation following a DBA. The licensee

did not propose changes to the actions or surveillance requirement associated with LCO 3.6.5.

Therefore, the addition of the proposed note is acceptable and would continue to meet 10 CFR 50.36.

3.8 Evaluation of Risk-Insights While this is not a risk-informed LAR, the licensee provided risk insights related to the proposed change in LAR Section 3, Technical Evaluation. The probabilistic risk assessment (PRA) models used by the licensee to derive risk insights were not reviewed by the NRC staff to determine their technical acceptability to support this safety evaluation.

The NRC staff considered the licensee-provided risk insights to aid the deterministic review of the proposed change and determined that the risk insights provided by the licensee support the traditional engineering conclusions associated with the proposed change.

Based on the information provided by the licensee, the NRC staff concludes that the submitted risk insights are acceptable for the purposes of supporting the deterministic evaluation.

3.9 Other Considerations The licensee presented compensatory measures, in its Enclosure to the LAR, that will be put in place during the periods where the containment average air temperature exceeds the current 120°F limit and stated that a review of several other miscellaneous events did not reveal any additional impacts from the proposed change. Below is a summary of the licensees compensatory measures described in section 2.5 of the

Enclosure:

Operation of all four containment air coolers on a fast speed with emergency SW aligned, Running containment recirculation fans in high speed and performing containment mini-purge continuously, Putting controls in place to ensure containment cooling system components and supporting systems are not removed from service, and Ensuring containment cooling systems are being protected.

The NRC staff reviewed the compensatory measures for completeness in understanding the licensees proposed action but does not rely on these compensatory measures as a basis for its approval.

3.10 Technical Evaluation Conclusion The NRC staff review of the proposed change considered the pressure and temperature conditions in the containment due to a spectrum of postulated LOCAs and secondary line breaks. The staff evaluated the technical justifications and the analyses presented by the licensee. The NRC staff finds the proposed emergency LAR to be acceptable based on the technical basis provided in the LAR, as supplemented, and the proposed short duration of the change requested.

Based upon the discussion in Sections 3.1 through 3.9 of this SE, the NRC staff concludes that with the proposed revision, the licensee will continue to meet the acceptance criteria for (1) primary containment functional design based on GDCs 16 and 50 for the containment and its associated systems being able to accommodate the calculated pressure and temperature

conditions resulting from any LOCA and (2) for GDC 38 for the containment heat removal system(s) to rapidly reduce the containment pressure and temperature to within their safe limits.

Based on its review, the NRC staff finds the request to approve a one-time NOTE to Appendix A TS LCO 3.6.5 to revise the limit on containment average air temperature from the existing value of 120 °F to new value of 122 °F is acceptable. The NRC staffs independent evaluation concluded that there is reasonable assurance that the proposed changes are bounded by the plant analysis of record and is, therefore, acceptable for the limited period of the one-time request.

4.0 EMERGENCY SITUATION The NRCs regulations in 10 CFR 50.91(a)(5) state that where the NRC finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plants licensed power level, the NRC may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the NRC will publish a notice of issuance under 10 CFR 2.106, providing for opportunity for a hearing and for public comment after issuance.

As discussed in SNCs submittal dated August 22, 2023 (ML23234A151), the licensee requested that the proposed amendments be reviewed by the NRC on an emergency basis.

The licensee stated that the emergency situation resulted from the unforeseen emergent conditions of the Farley site ambient temperature being persistently elevated, including the upcoming period of projected temperatures to exceed approximately 100°F for multiple days.

This ambient condition is projected by the licensee to result in both Farley Units 1 and 2 containment average air temperatures exceeding 120°F. Additionally, the licensee stated that the Southeastern Reliability Coordinator (SeRC) has elevated a Conservative System Operations (CSO) Watch to a CSO Warning, applicable to both Farley units.

The licensee further stated that without approval of this proposed change, a unit shutdown of one or both units could be required to commence (each Unit commencing when its respective containment average air temperature exceeds 120°F for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) and proceed until such time that containment average air temperature was restored to 120°F or Mode 5 (Cold Shutdown) is reached. Additionally, operation at a reduced power level is unlikely to result in containment average air temperature reductions to within the limit of TS 3.6.5 based on a review of containment temperature trends during unit down-power maneuvers.

The LCO 3.6.5 limit on containment average air temperature is proposed to be revised from 120°F to 122°F until 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on September 9, 2023. During this period, where the temperature limit of LCO 3.6.5 is temporarily increased, the licensee proposed certain compensatory measures. These measures are summarized above in section 3.9 of this SE.

NRC Staff Conclusion The NRC staff reviewed the licensees basis for processing the proposed amendments as an emergency amendment (as discussed above) and has determined that an emergency situation exists consistent with the provisions in 10 CFR 50.91(a)(5). Furthermore, the NRC staff determined that: (1) the licensee used its best efforts to make a timely application; (2) the licensee could not reasonably have avoided the situation; and (3) the licensee has not abused the provisions of 10 CFR 50.91(a)(5). Based on these findings, and the determination that the

amendments involve no significant hazards consideration as discussed below, the NRC staff has determined that a valid need exists for issuance of the license amendments using the emergency provisions of 10 CFR 50.91(a)(5).

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The licensees evaluation of the issue of no significant hazards consideration is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes do not adversely affect the operation of any structures, systems, or components (SSCs) associated with an accident initiator or initiating sequence of events. The proposed changes do not affect the design of the containment heat removal systems.

The proposed amendment does not affect accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility.

The proposed amendment does not alter any plant equipment or operating practices with respect to such initiators or precursors in a manner that the probability of an accident is increased. The proposed amendment to temporarily change the initial containment average air temperature does not adversely affect the operation of the assumed mitigation systems or the containment fission product barrier assumptions.

As demonstrated in the SNC request, the temporary increase in allowed containment temperature is more than offset by existing margins in the safety analyses. As such, the proposed temporary change will not alter assumptions relative to the mitigation of an accident or transient event. The proposed amendment does not increase the likelihood of the malfunction of an SSC or adversely impact analyzed accidents.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment does not introduce any new or unanalyzed modes of operation. The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the

methods governing normal plant operation. The changes do not alter the limiting assumptions made in the safety analysis.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The margin of safety is related to the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment. The performance of these fission product barriers is not affected by the proposed amendment; based on the pre-existing margins and conservatisms currently assumed in the safety analyses. Therefore, the margins to the onsite and offsite radiological dose limits are not significantly reduced.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendments and that the amendments should be issued as allowed by the criteria contained in 10 CFR 50.91.

6.0 STATE CONSULTATION

In accordance with the Commissions regulations, the State of Alabama official was notified of the proposed issuance of the amendments on August 22, 2023. The State official confirmed that the State of Alabama had no comments on August 23, 2023.

7.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facility component 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. Accordingly, the amendment meets 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 amendment.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) the amendments do not (a) involve a significant increase in the probability or consequences of an accident previously evaluated; or (b) create the possibility of a new or different kind of accident from any accident previously evaluated; or (c) involve a significant reduction in a margin of safety; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (3) there is reasonable assurance that such

activities will be conducted in compliance with the Commissions regulations, and (4) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Santosh Bhatt Joseph Messina Jay Robinson Matthew McConnell Steve Smith Dawnmathews T. Kalathiveettil Date: August 24, 2023

ML23235A296 NRR-058 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/SNSB/BC NAME DKalathiveettil KGoldstein PSahd DATE 8/23/2023 8/23/2023 8/23/2023 OFFICE NRR/DSS/SCPB/BC (A) NRR/DEX/ELTB/BC (A) NRR/DSS/STSB/BC (A)

NAME HWagage LRamadan MJardaneh DATE 8/23/2023 8/23/2023 8/23/2023 OFFICE NRR/DRA/APLB/BC OGC - NLO NRR/DORL/LPL2-1/BC NAME JWhitman MWoods MMarkley DATE 8/23/2023 8/24/2023 8/24/2023 OFFICE NRR/DORL/LPL2-1/PM NAME DKalathiveettil DATE 8/24/2023