Text

Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1
	ML23195A078
Person / Time
Site:	Oconee, Mcguire, Catawba, Harris, Brunswick, Robinson, McGuire
Issue date:	08/29/2023
From:	Shawn Williams; Plant Licensing Branch II
To:	Gibby S; Duke Energy Corp
	Jordan, N
References
	EPID L-2023-LLA-0028
	Download: ML23195A078 (80)
	v • d • e

August 29, 2023 Mr. Shawn Gibby Vice President Nuclear Engineering Duke Energy 526 South Church Street, EC-07H Charlotte, NC 28202

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2; CATAWBA NUCLEAR STATION, UNIT NOS. 1 AND 2; SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1; MCGUIRE NUCLEAR STATION, UNIT NOS. 1 AND 2; OCONEE NUCLEAR STATION, UNIT NOS. 1, 2, AND 3; AND H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENTS REGARDING THE ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER TSTF-554, REVISION 1 (EPID L-2023-LLA-0028)

Dear Mr. Gibby:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the following enclosed Amendment Nos. 312 and 340 to Renewed Facility Operating License Nos. DPR-71 and DPR-62 for the Brunswick Steam Electric Plant, Unit Nos. 1 and 2, respectively; Amendment Nos. 317 and 313 to Renewed Facility Operating License Nos. NPF-35 and NPF-52 for the Catawba Nuclear Station, Unit Nos. 1 and 2, respectively; Amendment No. 198 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit No. 1; Amendment Nos. 328 and 307 to Renewed Facility Operating License Nos. NPF-9 and NPF-17 for the McGuire Nuclear Station, Unit Nos. 1 and 2, respectively; Amendment Nos. 428, 430, and 429 to Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55 for the Oconee Nuclear Station, Unit Nos. 1, 2, and 3, (ONS) respectively; and Amendment No. 276 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2.

The license amendments consist of changes to the technical specifications (TSs) in response to your application dated February 16, 2023. The amendments revise the TSs related to reactor coolant system operational leakage and the definition of the term LEAKAGE based on Technical Specifications Task Force (TSTF) Traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements.

The issuance of the ONS Amendments also corrects an administrative error introduced on ONS TS page 1.1-4. The error was introduced in Amendment Nos. 367, 369, and 368 (ML101390415) on May 28, 2010, and inadvertently undid an approved change to TS page 1.1-4 in Amendment Nos. 366, 368, and 367 (Agencywide Documents Access and Management System Accession No. ML100220016) issued on January 28, 2010.

A copy of the NRC staffs safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.

If you have any questions, please contact me at (301) 415-1009 or by e-mail at Shawn.Williams@nrc.gov.

Sincerely,

/RA/

Shawn A. Williams, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.: 50-325, 50-324, 50-413, 50-414, 50-400, 50-369, 50-370, 50-269, 50-270, 50-287, and 50-261

Enclosures:

1. Amendment No. 312 to DPR-71(Brunswick 1)

2. Amendment No. 340 to DPR-62 (Brunswick 2)

3. Amendment No. 317 to NPF-35 (Catawba 1)

4. Amendment No. 313 to NPF-52 (Catawba 2)

5. Amendment No. 198 to DPR-63 (Harris 1)

6. Amendment No. 328 to NPF-9 (McGuire 1)

7. Amendment No. 307 to NPF-17 (McGuire 2)

8. Amendment No. 428 to DPR-38 (Oconee 1)

9. Amendment No. 430 to DPR-47 (Oconee 2)

10. Amendment No. 429 to DPR-55 (Oconee 3)

11. Amendment No. 276 to DPR-23 (Robinson 2)

12. Safety Evaluation cc: Listserv

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 312 Renewed License No. DPR-71

1.

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

A.

The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated February 16, 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.

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. DPR-71 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 312, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:08:36 -04'00'

ATTACHMENT TO DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT NO. 312 RENEWED FACILITY OPERATING LICENSE NO. DPR-71 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert DPR-71, Page 6 DPR-71, Page 6 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-4 1.1-4 1.1-5 1.1-5 3.4-7 3.4-7 3.4-8 3.4-8 Renewed License No. DPR-71 Amendment No. 312 (c)

Transition License Conditions

1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensees fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.

2. The licensee shall implement the modifications to its facility, as described in Table S-1, Plant Modifications Committed, of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3. The licensee shall complete all implementation items, except item 9, listed in LAR Attachment S, Table S-2, Implementation Items, of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 180th day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.

C.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 312, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 203 to Renewed Facility Operating License DPR-71, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 203. For SRs that existed prior to Amendment 203, including SRs with modified acceptance criteria and SRs whose frequency of

Definitions 1.1 Brunswick Unit 1 1.1-4 Amendment No. 312 1.1 Definitions DRAIN TIME d) No additional draining events occur; and (continued) e) Realistic cross-sectional areas and drain rates are used.

A bounding DRAIN TIME may be used in lieu of a calculated value.

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval from SYSTEM (ECCS) RESPONSE when the monitored parameter exceeds its ECCS initiation TIME setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

ISOLATION INSTRUMENTATION The ISOLATION INSTRUMENTATION RESPONSE TIME RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves receive the isolation signal (e.g.,

de-energization of the MSIV solenoids). The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or 2.

LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; (continued)

Definitions 1.1 Brunswick Unit 1 1.1-5 Amendment No. 312 1.1 Definitions LEAKAGE b.

Unidentified LEAKAGE (continued)

All LEAKAGE into the drywell that is not identified LEAKAGE c.

Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and d.

Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

LINEAR HEAT The LHGR shall be the heat generation rate per unit length of GENERATION RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of TEST all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR)

RATIO (MCPR) that exists in the core. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(continued)

RCS Operational LEAKAGE 3.4.4 Brunswick Unit 1 3.4-7 Amendment No. 312 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE; b.

5 gpm unidentified LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; c.

25 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; and d.

2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in MODE 1.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RCS Operational LEAKAGE 3.4.4 Brunswick Unit 1 3.4-8 Amendment No. 312 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Unidentified LEAKAGE not within limit.

OR Total LEAKAGE not within limit.

OR Unidentified LEAKAGE increase not within limit.

B.1 Reduce LEAKAGE to within limits.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C.

Required Action and associated Completion Time not met.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and unidentified LEAKAGE increase are within limits.

In accordance with the Surveillance Frequency Control Program

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 340 Renewed License No. DPR-62

1.

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

A.

The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated February 16, 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.

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. DPR-62 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 340, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:09:05 -04'00'

ATTACHMENT TO DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT NO. 340 RENEWED FACILITY OPERATING LICENSE NO. DPR-62 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert DPR-62, Page 6 DPR-62, Page 6 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-4 1.1-4 1.1-5 1.1-5 3.4-7 3.4-7 3.4-8 3.4-8 Renewed License No. DPR-62 Amendment No. 340 (c)

Transition License Conditions

1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensees fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.

2. The licensee shall implement the modifications to its facility, as described in Table S-1, Plant Modifications Committed, of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3. The licensee shall complete all implementation items, except item 9, listed in LAR Attachment S, Table S-2, Implementation Items, of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 180th day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.

C.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(3)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts thermal.

(4)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 340, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 203 to Renewed Facility Operating License DPR-71, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 203. For SRs that existed prior to Amendment 203, including SRs with modified acceptance criteria and SRs whose frequency of

Definitions 1.1 Brunswick Unit 2 1.1-4 Amendment No. 340 1.1 Definitions DRAIN TIME d) No additional draining events occur; and (continued) e) Realistic cross-sectional areas and drain rates are used.

A bounding DRAIN TIME may be used in lieu of a calculated value.

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

ISOLATION INSTRUMENTATION The ISOLATION INSTRUMENTATION RESPONSE TIME RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves receive the isolation signal (e.g.,

de-energization of the MSIV solenoids). The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or 2.

LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; (continued)

Definitions 1.1 Brunswick Unit 2 1.1-5 Amendment No. 340 1.1 Definitions LEAKAGE b.

Unidentified LEAKAGE (continued)

All LEAKAGE into the drywell that is not identified LEAKAGE; c.

Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and d.

Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

LINEAR HEAT The LHGR shall be the heat generation rate per unit length of GENERATION RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR)

RATIO (MCPR) that exists in the core. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(continued)

RCS Operational LEAKAGE 3.4.4 Brunswick Unit 2 3.4-7 Amendment No. 340 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE; b.

5 gpm unidentified LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; c.

25 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; and d.

2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in MODE 1.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RCS Operational LEAKAGE 3.4.4 Brunswick Unit 2 3.4-8 Amendment No. 340 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Unidentified LEAKAGE not within limit.

OR Total LEAKAGE not within limit.

OR Unidentified LEAKAGE increase not within limit.

B.1 Reduce LEAKAGE to within limits.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C.

Required Action and associated Completion Time not met.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and unidentified LEAKAGE increase are within limits.

In accordance with the Surveillance Frequency Control Program

DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 317 Renewed License No. NPF-35

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by Duke Energy Carolinas, LLC (the licensee), dated February 16, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.

2.

Accordingly, the license is hereby amended by page 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-35 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 317, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:09:57 -04'00'

DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 313 Renewed License No. NPF-52

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by the Duke Energy Carolinas, LLC (the licensee), dated February 16, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.

2.

Accordingly, the license is hereby amended by page 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-52 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 313, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC, shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:10:24 -04'00'

ATTACHMENT TO DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT NO. 317 RENEWED FACILITY OPERATING LICENSE NO NPF-35 AND DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT NO. 313 RENEWED FACILITY OPERATING LICENSE NO. NPF-52 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert NPF-35, Page 4 NPF-35, Page 4 NPF-52, Page 4 NPF-52, Page 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-4 1.1-4 3.4.13-1 3.4.13-1 3.4.13-2 3.4.13-2 N/A 3.4.13-3 Renewed License No. NPF-35 Amendment No. 317 (2)

TECHNICAL SPECIFICATIONS The Technical Specifications contained in Appendix A, as revised through Amendment No. 317, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3)

Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4)

Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5)

Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. NPF-525 Amendment No. 313 (2)

TECHNICAL SPECIFICATIONS The Technical Specifications contained in Appendix A, as revised through Amendment No. 313, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3)

Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4),

following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4)

Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5)

Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013, as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c),

the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Definitions 1.1 1.1 Definitions (continued)

Catawba Units 1 and 2 1.1-4 (continued)

Amendment Nos. 317/313 LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or 3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

RCS Operational LEAKAGE 3.4.13 Catawba Units 1 and 2 3.4.13-1 Amendment Nos. 317/313 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gallons per day (Unit 1) and 45 gallons per day (Unit 2) primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RCS Operational LEAKAGE 3.4.13 Catawba Units 1 and 2 3.4.13-2 Amendment Nos. 317/313 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and associated Completion Time not met.

OR Primary to secondary LEAKAGE not within limit.

C.1 Be in MODE 3.

AND C.2 Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours

RCS Operational LEAKAGE 3.4.13 Catawba Units 1 and 2 3.4.13-3 Amendment Nos. 317/313 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ------------------------------NOTES----------------------------------

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS Operational LEAKAGE within limits by performance of RCS water inventory balance.

NOTE------

Only required to be performed during steady state operation In accordance with the Surveillance Frequency Control Program SR 3.4.13.2 ----------------------------------NOTE---------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is < 150 gallons per day (Unit 1) and < 45 gallons per day (Unit 2) through any one SG.

NOTE--------

Only required to be performed during steady state operation In accordance with the Surveillance Frequency Control Program

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 198 Renewed License No. NPF-63

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Progress, LLC (the licensee),

dated February 16, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

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

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

2.

Accordingly, the license is amended by 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-63 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 198, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating License and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:10:53 -04'00'

ATTACHMENT TO DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT NO. 198 RENEWED FACILITY OPERATING LICENSE NO. NPF-63 Replace the following page of the Renewed Facility Operating License with the revised page.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert NPF-63, Page 4 NPF-63, Page 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1-3 1-3 1-4 1-4 3/4 4-23 3/4 4-23 Renewed License No. NPF-63 Amendment No. 198 C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's 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 incorporated below.

(1)

Maximum Power Level Duke Energy Progress, LLC, is authorized to operate the facility at reactor Core power levels not in excess of 2948 megawatts thermal (100 percent rated core power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 198, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Antitrust Conditions Duke Energy Progress, LLC. shall comply with the antitrust conditions delineated in Appendix C to this license.

(4)

Initial Startup Test Program (Section 14)1 Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

(5)

Steam Generator Tube Rupture (Section 15.6.3)

Prior to startup following the first refueling outage, Carolina Power & Light Company* shall submit for NRC review and receive approval if a steam generator tube rupture analysis, including the assumed operator actions, which demonstrates that the consequences of the design basis steam generator tube rupture event for the Shearon Harris Nuclear Power Plant are less than the acceptance criteria specified in the Standard Review Plan, NUREG-0800, at 15.6.3 Subparts II (1) and (2) for calculated doses from radiological releases. In preparing their analysis Carolina Power &

Light Company* will not assume that operators will complete corrective actions within the first thirty minutes after a steam generator tube rupture.

1The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

On April 29, 2013, the name of Carolina Power & Light Company (CP&L) was changed to Duke Energy Progress, Inc. On August 1, 2015, the name Duke Energy Progress, Inc. was changed to Duke Energy Progress, LLC.

SHEARON HARRIS - UNIT 1 1-3 Amendment No. 198 DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration (MeV/d) for isotopes, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

EXCLUSION AREA BOUNDARY 1.14 The EXCLUSION AREA BOUNDARY shall be that line beyond which the land is not controlled by the licensee to limit access.

FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

1.16 (DELETED)

IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:

a.

Leakage, such as that from pump seals or valve packing (except CONTROLLED LEAKAGE), that is captured and conducted to a sump or collecting tank, or

b.

Leakage into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of Leakage Detection Systems; or

c.

Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary to secondary leakage).

INSERVICE TESTING PROGRAM 1.17a The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

SHEARON HARRIS - UNIT 1 1-4 Amendment No. 198 DEFINITIONS MASTER RELAY TEST 1.18 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER(S) OF THE PUBLIC 1.19 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.20 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.3 and 6.9.1.4.

OPERABLE - OPERABILITY 1.21 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.22 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.23 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation:

(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.24 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a fault in a Reactor Coolant System component body, pipe wall, or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

Test pressures less than 2235 psig but greater than 150 psig are allowed. Observed leakage shall be adjusted by multiplying the observed leakage by the square root of the quotient of 2235 divided by the test pressure.

SHEARON HARRIS - UNIT 1 3/4 4-23 Amendment No. 198 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 gpm UNIDENTIFIED LEAKAGE, c.

150 gallons per day primary to secondary leakage through any one steam generator, d.

10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, e.

31 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 +/- 20 psig, and f.

The maximum allowable leakage of any Reactor Coolant System Pressure Isolation Valve shall be as specified in Table 3.4-1 at a pressure of 2235 +/- 20 psig.*

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

ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE, isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding primary to secondary leakage, PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit specified in Table 3.4-1, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d.

With primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 328 Renewed License No. NPF-9

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Carolinas, LLC (the licensee),

dated February 16, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

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

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

2.

Accordingly, the license is hereby amended by page 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-9 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 328, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:11:22 -04'00'

DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 307 Renewed License No. NPF-17

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Duke Energy Carolinas, LLC (the licensee), dated February 16, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 set forth in 10 CFR Chapter I; D.

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

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

2.

Accordingly, the license is hereby amended by page 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-17 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 307, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:11:47 -04'00'

ATTACHMENT TO DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT NO. 328 RENEWED FACILITY OPERATING LICENSE NO. NPF-9 AND DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT NO. 307 RENEWED FACILITY OPERATING LICENSE NO. NPF-17 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert NPF-9, Page 3 NPF-9, Page 3 NPF-17, Page 3 NPF-17, Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-3 1.1-3 3.4.13-1 3.4.13-1 3.4.13-2 3.4.13-2 N/A 3.4.13-3

Renewed License No. NPF-9 Amendment No. 328 (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, 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; (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and; (6)

Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's 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 incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%).

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 328, are hereby incorporated into this renewed operating license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(3)

Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.

Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. NPF-17 Amendment No. 307 (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, 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; (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and; (6)

Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's 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 incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%).

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 307, are hereby incorporated into this renewed operating license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(3)

Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than March 3, 2023, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.

Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Definitions 1.1 1.1 Definitions (continued)

(continued)

McGuire Units 1 and 2 1.1-3 Amendment Nos. 328/307 ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its ESF actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and

c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

RCS Operational LEAKAGE 3.4.13 McGuire Units 1 and 2 3.4.13-1 Amendment No. 328/307 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE;

d.

389 gallons per day total primary to secondary LEAKAGE through all steam generators (SGs); and

e.

135 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

RCS Operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RCS Operational LEAKAGE 3.4.13 McGuire Units 1 and 2 3.4.13-2 Amendment No. 328/307 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and associated Completion Time not met.

OR Primary to secondary LEAKAGE not within limits.

C.1 Be in MODE 3.

AND C.2 Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours

RCS Operational LEAKAGE 3.4.13 McGuire Units 1 and 2 3.4.13-3 Amendment No. 328/307 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ------------------------------NOTES----------------------------------

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS Operational LEAKAGE is within limits by performance of RCS water inventory balance.

NOTE-------

Only required to be performed during steady state operation In accordance with the Surveillance Frequency Control Program SR 3.4.13.2 -------------------------------NOTE----------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is < 135 gallons per day through any one SG and < 389 gallons per day total through all SGs.

In accordance with the Surveillance Frequency Control Program

DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 428 Renewed License No. DPR-38

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Carolinas, LLC (the licensee),

dated February 16, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 set forth in 10 CFR Chapter I; D.

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

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

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 428 are hereby incorporated into the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:12:19 -04'00'

DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 430 Renewed License No. DPR-47

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Carolinas, LLC (the licensee),

dated February 16, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 430 are hereby incorporated into the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:12:43 -04'00'

0 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 429 Renewed License No. DPR-55

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Carolinas, LLC (the licensee),

dated February 16, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 429 are hereby incorporated into the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating Licenses and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:13:19 -04'00'

ATTACHMENT TO DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT NO. 428 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 AND DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT NO. 430 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 AND DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT NO. 429 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert DPR-38, Page 3 DPR-38, Page 3 DPR-47, Page 3 DPR-47, Page 3 DPR-55, Page 3 DPR-55, Page 3 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-4 1.1-4 3.4.13-1 3.4.13-1 3.4-13-2 3.4.13-2 Renewed License No. DPR-38 Amendment No. 428 A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 428 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1.

As used herein:

(a)

Bulk Power means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b)

Neighboring Entity means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-47 Amendment No. 430 A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 430 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

2.

As used herein:

(b)

"Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(c)

"Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-55 Amendment No. 429 A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 429 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

3.

As used herein:

(d)

"Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(e)

"Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of

Definitions 1.1 1.1 Definitions (continued)

OCONEE UNITS 1, 2, & 3 1.1-4 Amendment Nos. 428, 430 & 429 LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or 3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

RCS Operational LEAKAGE 3.4.13 OCONEE UNITS 1, 2, & 3 3.4.13-1 Amendment Nos. 428, 430 & 429 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE; b.

1 gpm unidentified LEAKAGE; c.

10 gpm identified LEAKAGE; and d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)

RCS Operational LEAKAGE 3.4.13 OCONEE UNITS 1, 2, & 3 3.4.13-2 Amendment Nos. 428, 430 & 429 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and associated Completion Time not met.

OR Primary to secondary LEAKAGE not within limit.

C.1 Be in MODE 3.

AND C.2 Be in MODE 5.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

1 DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 Amendment No. 276 Renewed License No. DPR-23

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Progress, LLC (the licensee),

dated February 16, 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 Commission's regulations set forth in 10 CFR Chapter I; 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.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 3.B. of Renewed Facility Operating License No. DPR-23 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

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 Renewed Facility Operating License and Technical Specifications Date of Issuance: August 29, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.08.29 13:13:58 -04'00'

ATTACHMENT TO DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT UNIT 2 AMENDMENT NO. 276 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert DPR-23, Page 3 DPR-23, Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-3 1.1-3 3.4-35 3.4-35 Renewed Facility Operating License No. DPR-23 Amendment No. 276 D.

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; E.

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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 incorporated below:

A.

Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(1)

For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.

Definitions 1.1 1.1 Definitions (continued)

HBRSEP Unit No. 2 1.1-3 Amendment No. 276

- AVERAGE iodines, with half lives > 15 minutes, making up DISINTEGRATION ENERGY at least 95% of the total noniodine activity in (continued) the coolant.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or return), that is captured and conducted to collection systems or a sump or collecting tank; 2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or 3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or return) that is not identified LEAKAGE; and c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

RCS Operational LEAKAGE 3.4.13 HBRSEP Unit No. 2 3.4-35 Amendment No. 276 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE; b.

1 gpm unidentified LEAKAGE; c.

10 gpm identified LEAKAGE; and d.

75 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C.

Required Action and associated Completion Time not met.

OR Primary to secondary LEAKAGE not within limit.

C.1 Be in MODE 3.

AND C.2 Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours

2 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TECHNICAL SPECIFICATIONS TASK FORCE (TSTF) TRAVELER TSTF-554, REVISION 1 REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS DUKE ENERGY CAROLINAS, LLC AND DUKE ENERGY PROGRESS, LLC BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 CATAWBA NUCLEAR STATION, UNIT NOS. 1 AND 2 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 MCGUIRE NUCLEAR STATION, UNIT NOS. 1 AND 2 OCONEE NUCLEAR STATION, UNIT NOS. 1, 2, AND 3 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NOS. 50-325, 50-324, 50-413, 50-414, 50-400, 50-369, 50-370, 50-269, 50-270, 50-287, AND 50-261

1.0 INTRODUCTION

By application dated February 16, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23047A004), Duke Energy Progress, LLC and Duke Energy Carolinas, LLC, collectively referred to as Duke Energy, requested amendments to the Technical Specifications (TSs) for the Brunswick Steam Electric Plant, Unit Nos. 1 and 2 (BNP);

the Catawba Nuclear Station, Unit Nos. 1 and 2 (CNS); the Shearon Harris Nuclear Power Plant, Unit No. 1 (HNP); the McGuire Nuclear Station, Unit Nos. 1 and 2 (MNS); the Oconee Nuclear Station, Unit Nos. 1, 2, and 3 (ONS); and the H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP).

Duke Energy requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) process the proposed LAR under the Consolidated Line Item Improvement Process (CLIIP).

The proposed changes would revise the TSs related to reactor coolant system (RCS) operational leakage and the definition of the term LEAKAGE based on Technical Specifications Task Force (TSTF) Traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements, (TSTF-554) (ML20016A233). The NRC issued the final safety evaluation approving the use of TSTF-554, Revision 1, on December 18, 2020 (ML20322A024).

Duke Energy proposed variations from the TS changes described in Traveler TSTF-554, Revision 1. The variations are described section 2.2, Enclosure 1 of the license amendment request (LAR), and evaluated in section 3.2 of this SE.

1.1 Reactor Coolant System Description Components that contain or transport the coolant to or from the reactor core make up the RCS.

Materials can degrade as a result of the complex interaction of the materials, the stresses they encounter, and through operational wear or mechanical deterioration during normal and upset operating environments. Such material degradation could lead to leakage of reactor coolant into containment buildings.

The RCS leakage falls under two main categories - identified leakage and unidentified leakage.

Identifying the sources of leakage is necessary for prompt identification of potentially adverse conditions, assessment of safety significance of the leakage, and quick corrective action. A limited amount of leakage from the reactor coolant pressure boundary (RCPB) directly into the containment atmosphere is expected as the RCS and other connected systems cannot be made 100 percent leak tight. This leakage is detected, located, and isolated from the containment atmosphere so as to not interfere with measurement of unexpected RCS leakage detection.

The safety significance of RCS leakage varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS leakage into the containment area is necessary. Separation of identified leakage from unidentified leakage provides quantitative information to the operators, allowing them to take corrective action should leakage occur that is detrimental to the safety of the unit and the public.

1.2 Proposed TS Changes to Adopt TSTF-554 The licensee proposed changes that would revise the TSs related to RCS operational leakage and the definition of the term LEAKAGE. Specifically, the licensee proposed the following changes be made to the Duke Fleet TSs in order to adopt TSTF-554:

The identified LEAKAGE definition in TS 1.1.a.2 for BNP Units 1 and 2, CNS Units 1 and 2, MNS Units 1 and 2, ONS Units 1, 2 and 3, RNP Unit 2 and TS 1.17a for HNP Unit 1, would be revised to remove the exclusion of pressure boundary leakage from identified leakage by deleting either and the phrase not to be pressure boundary LEAKAGE.

The pressure boundary LEAKAGE definition in TS 1.1.d for BNP Units 1 and 2, TS 1.1.c for CNS Units 1 and 2, MNS Units 1 and 2, ONS Units 1, 2 and 3, RNP Unit 2 and TS 1.24 for HNP Unit 1, would be revised to delete the word nonisolable. The sentence, LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE, would be adopted from the Standard TS (STS) Bases and added to the definition.

Additionally, for each facility, the LEAKAGE definition would be revised by other editorial and punctuation changes to reflect the deletion and listed definitions.

The TS RCS Operational LEAKAGE ACTIONS would be revised (a) to add a new Condition to address isolation of the pressure boundary leakage within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , (b) to be applicable should the TS ACTION of new condition is not met, and (c) to delete the condition for when pressure boundary leakage exists because pressure boundary leakage would be addressed by the new Condition.

Plant specific TS changes (shown in bold text) are as follows:

Changes to TS 3.4.6.2 for HNP Unit 1:

o Current ACTION a:

With any PRESSURE BOUNDARY LEAKAGE or with primary-to-secondary leakage not within limit. be in at least HOT STANDBY within b. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Revised ACTION a:

With any PRESSURE BOUNDARY LEAKAGE, or with primary-to-secondary leakage not within limit isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

o No proposed changes to current ACTIONS b and c.

o Current ACTION a above, specifies requirements for condition when the primary-to-secondary leakage is not within limit. This condition is renumbered as new ACTION d as below, with no changes to the current ACTION as Required Actions and Completion Time:

ACTION d states, With primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Proposed changes to Conditions and Required Actions [existing completion times are not affected in the current TS] for TS 3.4.4 for BNP Units 1 and 2, TS 3.4.13 for CNS Units 1 and 2, MNS Units 1 and 2, ONS Units 1, 2, 3 and RNP Unit 2 as follows:

CONDITION REQUIRED ACTION COMPLETION TIME New CONDITION A Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Current CONDITION A renumbered as CONDITION B with no changes in Required Actions and Completion Times.

Current CONDITION B renumbered as CONDITION C with the following changes.

CONDITION B C B C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action and associated Completion Time of Condition A not met.

OR Pressure boundary LEAKAGE exists.

OR (N/A for BNP Units)

Primary to secondary LEAKAGE not within limit.

AND B C.2 Be in MODE 5.

Be in MODE 4 (BNP) 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 1.3 Additional Proposed TS Changes The application identified certain variations from TSTF-554. Section 3.2 of this SE provides a description of the variations and the NRC staffs evaluation.

2.0 REGULATORY EVALUATION

The regulation at Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(2) requires that TSs include limiting conditions for operation (LCOs). Per 10 CFR 50.36(c)(2)(i), LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulation also requires that [w]hen a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The regulation at 10 CFR 50.2, Definitions, defines the reactor coolant pressure boundary as all those pressure-containing components of boiling and pressurized water-cooled nuclear power reactors, such as pressure vessels, piping, pumps, and valves.

Regulatory Guide (RG) 1.45, Revision 1, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, dated May 1, 2008 (ML073200271), Section B, Discussion Leakage Separation, provides information related to separation between identified and unidentified leakage.

The NRC staffs guidance for the review of TSs is in Chapter 16.0, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), March 2010 (ML100351425).

As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREGs 14301, 1U.S. Nuclear Regulatory Commission Standard Technical Specifications, Babcock and Wilcox, NUREG-1430, Volume 1, Specifications, and Volume 2, Bases, Revision 5.0, dated September 2021 (ML21272A363 and ML21272A370, respectively).

14312, and 14333 as modified by NRC approved travelers. Traveler TSTF-554 revised the STSs related to RCS operational leakage and the definition of the term LEAKAGE. The NRC approved TSTF-554, under the CLIIP on December 18, 2020.

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes to Adopt TSTF-554 The NRC staff compared the licensees proposed TS changes in Section 1.1 of this SE against the changes approved in TSTF-554. In accordance with the SRP Chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-554 are applicable to the Duke Energy units (as listed in the application) because these facilities are pressurized water reactors (PWRs) and boiling water reactors (BWR), and the NRC staff approved the TSTF-554 changes for PWR as well as the BWR design. The NRC finds that the licensees proposed changes to the TS definitions sections as listed in section 1.2 of this safety evaluation are consistent with those found acceptable in TSTF-554.

In the SE of TSTF-554, the NRC staff concluded that TSTF-554 changes to the STS 1.1 definition of LEAKAGE and to the STS 3.4.4 and 3.4.13, the LCOs addressing conditions and required actions when RCS pressure boundary leakage exists, are acceptable. The NRC staff found that removing the term nonisolable provides a clearer definition of pressure boundary leakage and that the source of the leakage is not relevant to this capability provided that separate, appropriate limits on pressure boundary leakage have been established. Therefore, the proposed change to the definition of identified leakage was acceptable as it did not conflict with the RCPB definition in 10 CFR 50.2 and was consistent with RG 1.45. The NRC staff further found that proposed new Condition A in TSs 3.4.4, 3.4.13 and ACTION a in TS 3.4.6.2 on boundary pressure leakage, including its associated and Completion Time, acceptable because the LCO revisions continue to specify the lowest functionable capability of equipment, identify remedial actions and require shutdown of the reactor if the remedial actions cannot be met.

The NRC staff finds that proposed changes to the TS 1.1 definition for BNP Units 1 and 2, CNS Units 1 and 2, MNS Units 1 and 2, ONS Units 1, 2 and 3, and RNP Unit 2, and TS 1.17 and 1.24 for HNP Unit 1 clarify what constitutes pressure boundary leakage and the source of leakage does not matter if the TSs have separate limits on pressure boundary leakage and LCOs 3.4.4, for BNP Units 1 and 2, TS 4.4.13 for CNS Units 1 and 2, MNS Units 1 and 2, ONS Units 1, 2 and 3, and RNP Unit 2, and 3.4.6.2 for HNP Unit 1, correctly specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. Also, the NRC staff finds that proposed changes to the LCOs Actions are adequate remedial actions to be taken until each LCO can be met provide protection to the health and safety of the public.

Thus, the proposed changes continue to meet the requirements of 10 CFR 50.36(c)(2)(i) as discussed in section 3.0 of the NRC staffs SE of TSTF-554.

²U.S. Nuclear Regulatory Commission Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 5.0, dated September 2021 (ML21259A155 and ML21259A159, respectively).

³U.S. Nuclear Regulatory Commission Standard Technical Specifications, General Electric BWR/4 Plants, NUREG-1433, Volume 1, Specifications, and Volume 2, Bases, Revision 5.0, dated September 2021 (ML21272A357 and ML21272A358, respectively).

3.2 Technical Evaluation of Additional Proposed TS Changes Duke Energy is proposing the following variation from the TS changes described in TSTF-554 or the applicable parts of the NRC staff safety evaluation, dated December 18, 2020:

BNP The licensee states:

The affected portions of the BNP TS have some editorial differences from the corresponding STS on which TSTF-554 was based. None of the differences affect the applicability of TSTF-554 to the BNP TS.

BNP TS 1.1 Definition for Identified LEAKAGE, Item a.1, already ends with a ; so the markups in this license amendment request differ editorially from TSTF-554 which shows the punctuation changing from a. to a ;.

BNP TS 1.1 Definition for Identified LEAKAGE Item a.2, already ends with a ; so the markups in this license amendment request differ editorially from TSTF-554 which shows the punctuation changing from a. to a ;.

BNP TS 1.1 Definition for Unidentified LEAKAGE, Item b, ends with a ; so the markups in this license amendment request differ editorially from TSTF-554 which shows the punctuation ending with a BNP TS 1.1 Definition for Total LEAKAGE, Item c, already contains a ; prior to the and so the markups in this license amendment request differ editorially from TSTF-554 which shows the punctuation changing from a, to a ;.

The licensee further states:

The BNP TS 3.4.4, RCS Operational LEAKAGE, Conditions are structured differently than those in the Standard Technical Specifications (STS) on which TSTF-554 were based. Specifically, the following difference exists.

The existing STS 3.4.4. contains Condition A, which addresses unidentified LEAKAGE not within limit or total LEAKAGE not within limit and Condition B, which addresses increase in unidentified LEAKAGE not within limit. The existing BNP TS 3.4.4, Condition A addresses unidentified LEAKAGE not within limit, total LEAKAGE not within limit, or increase in unidentified LEAKAGE not within limit.

The structural difference (i.e., existing BNP Condition A versus existing STS Conditions A and B) is administrative and does not affect the applicability of TSTF-554, Revision 1, to the BNP TS. The requirements of existing BNP TS 3.4.4 differ from those in the STS on which TSTF-554 were based. Specifically, the following differences exist.

The existing STS 3.4.4. Conditions A and B have a 4-hour Completion Time. The existing equivalent BNP Condition A has an 8-hour Completion Time.

Existing STS 3.4.4, Required Action B.2 is not included in the BNP TS 3.4.4 Required Actions.

TSTF-554, Revision 1, establishes a new TS 3.4.4 Condition A to address pressure boundary LEAKAGE with a Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the 4-hour Completion time cannot be met, the plant must initiate a shutdown (i.e., be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Mode 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ). The proposed BNP revision incorporates the same requirements as TSTF-554, Revision 1. The 8-hour Completion Time of existing BNP Required Action A.1 (i.e., new Required Action B.1) has no impact on the applicability of TSTF-554, Revision 1, to the BNP TS.

STS 3.4.4, Required Action B.2 allows an option to verify the source of unidentified LEAKAGE is not service sensitive type 304 or type 316 austenitic stainless steel. The lack of a BNP Required Action that is equivalent to STS 3.4.4, Required Action B.2 results in the BNP TS 3.4.4 being more conservative than STS 3.4.4; any increase in unidentified LEAKAGE must be reduced to within limits independent of its source.

This difference has no impact on the applicability of TSTF-554, Revision 1, to the BNP TS.

CNS The licensee states:

The affected portions of the CNS TS have some editorial differences from the corresponding STS on which TSTF-554 was based. None of the differences affect the applicability of TSTF-554 to the CNS TS.

CNS TS 1.1 Definition for Identified LEAKAGE, Item a.3, includes the acronym SG for steam generator while the STS does not. The acronym has been deleted as it is no longer needed based on the subsequent changes to align wording in the definition with TS 3.4.13.

CNS TS 1.1 Definition for Identified LEAKAGE, Item a.3 does not contain the phrase primary to secondary LEAKAGE in parentheses consistent with the STS. The parenthetical has been added to ensure consistent phrasing between the LEAKAGE definition and TS 3.4.13.

CNS TS 1.1 Definition for Identified LEAKAGE, Item a.3, already ends with a ; so the markups in this license amendment request differ editorially from TSTF-554 which shows the punctuation changing from a. to a ;.

CNS TS 1.1 Definition for Unidentified LEAKAGE Item b, does not end with and as shown in TSTF-554 and the STS. The and has been added for consistency with the STS.

CNS TS 1.1 Definition for Pressure Boundary LEAKAGE, Item c, uses the term SG LEAKAGE rather than primary to secondary LEAKAGE as shown in TSTF-554 and the STS. This license amendment request is requesting the editorial change to the definition to ensure consistent phrasing between the LEAKAGE definition and TS 3.4.13.

The licensee further states:

CNS TS 3.4.13 RCS Leakage LCO item d. is different from the STS on which TSTF-554 was based. The LCO provides a permanently reduced primary to secondary leakage rate for Unit 2 as 45 gallons per day. This was approved in Amendment 263 to the Unit 2 renewed facility operating license (ML12054A692). This limit was set based on evaluations performed after identification of stress corrosion cracking in the Unit 2 steam generators. This difference does not affect the applicability of TSTF-554 to the CNS TS.

HNP The licensee states:

The HNP TS are based upon the format and content of the NUREG-0452, Standard Technical Specifications for Westinghouse Pressurized Water Reactors, series. As a result, the HNP TS numbers, wording, and format varies from the NRC Improved Standard Technical Specifications shown in TSTF-554, Revision 1, and referenced in the applicable parts of the NRCs Safety Evaluation. These differences are administrative in nature and do not affect the applicability of TSTF-554 to the HNP TS.

The HNP TS 1.0, Definitions, does not contain a single definition for LEAKAGE. Instead, specific leakage type definitions are provided.

Table 1 shows a comparison of the current leakage definitions with the proposed definitions based on TSF-554.

TABLE 1 HNP TS Term Current Definition Proposed Definition 1.8 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

1.17 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or IDENTIFIED LEAKAGE shall be:

a. Leakage, such as that from pump seals or valve packing (except CONTROLLED LEAKAGE),

that is captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary-to-secondary leakage).

b. Leakage into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

c. Reactor Coolant System leakage through a steam generator to the Secondary System (primary to secondary leakage).

1.24 PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a fault in a Reactor Coolant System component body, pipe wall, or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE 1.39 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

The licensee further states:

As shown in Enclosure 2, the HNP TS 1.0 numbers and terms shown in Table 1 are maintained, while the definitions are modified to be consistent with TSTF-554. This is to ensure consistent numbering and terminology is maintained within the TS. The changes in this license amendment do not propose modification to the definition of CONTROLLED LEAKAGE. This term is used within the HNP TS but is not incorporated into the NUREG-1431 STS for Westinghouse plants, nor is the definition provided by TSTF-554, Revision 1. It is included in Table 1 to ensure clarity with the other definitions of leakage. The term UNIDENTIFIED LEAKAGE maintains the current TS wording as well since it refers to CONTROLLED LEAKAGE. These differences are administrative and do not affect the applicability of TSTF-554 to the HNP TS.

HNP TS 3.4.6.2, Reactor Coolant System Operational Leakage, contains additional LCO requirements not incorporated in STS or TSTF-554, Revision 1. The additional LCO requirements are not impacted by the application of TSTF-554, Revision 1 to the HNP TS.

HNP TS 3.4.6.2 Actions refer to HOT STANDBY and COLD SHUTDOWN operational modes instead of referring to the numerical MODES that are utilized in STS and TSTF-554, Revision 1. HNP TS 1.0, Table 1.2, Operational MODES, defines HOT STANDBY as MODE 3 and COLD SHUTDOWN as MODE 5. These differences are administrative and do not affect the applicability of TSTF-554 to the HNP TS.

HNP TS 3.4.6.2 LCO c. and Action b. contain an editorial difference from the corresponding STS on which TSTF-554 was based. The term primary-to-secondary is separated by hyphens (-) instead of spaces. As shown in Enclosure 2, the term primary to secondary no longer contains hyphens to ensure consistency with TSTF-554. This difference is administrative and does not affect the applicability of TSTF-554 to the HNP TS.

Since the HNP TS are based upon the format and content of the NUREG-0452, TS 3.4.6.2 Action a requires the Unit being at least in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months for any PRESSURE BOUNDARY LEAKAGE or primary to secondary leakage. To align with the wording and actions described in TSTF-554, this license amendment proposes revising TS 3.4.6.2 Action a to refer only to PRESSURE BOUNDARY LEAKAGE.

In addition, this license amendment includes a proposed change to create a new TS 3.4.6.2 Action d to describe actions for primary to secondary leakage not within the limit. As shown in Enclosure 2, with primary to secondary leakage not within the limit, the Unit must be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This change ensures consistency with TSTF-554 and the current HNP TS 3.4.6.2 requirements for primary to secondary leakage.

This difference is administrative and does not affect the applicability of TSTF-554 to the HNP TS. As a result, the HNP TS numbers, wording, and format varies from that which is included in STS and TSTF-554, Revision 1.

MNS The licensee states:

The affected portions of the MNS TS have some editorial differences from the corresponding STS on which TSTF-554 was based. None of the differences affect the applicability of TSTF-554 to the MNS TS.

MNS TS 1.1 Definition for Identified LEAKAGE Item a.3, already ends with a ; so the markups in this license amendment request differ editorially from TSTF-554 which shows the punctuation changing from a. to a ;.

MNS TS 1.1 Definition for Unidentified LEAKAGE, Item b, does not end with and as shown in TSTF-554 and the STS. The and has been added for consistency with the STS.

ONS The licensee states:

TSTF-554, Revision 1, and its corresponding Safety Evaluation discuss the applicable regulatory requirements and guidance, including the applicability of 10 CFR [Part] 50, Appendix A, General Design Criteria (GDC) 14 and 30. ONS Units 1, 2, and 3 were not licensed to the current 10 CFR [Part] 50, Appendix A, GDC. ONS Updated Final Safety Analysis Report (UFSAR) Chapter 3, Conformance with NRC General Design Criteria, states ONS was evaluated with respect to the proposed GDC contained in the Federal Register notice published July 11, 1967 (ML043310029). The UFSAR contains a discussion of the criteria as well as a summary of the criteria by groups. The ONS UFSAR provides criteria for the Reactor Coolant Pressure Boundary in Section 3.1.9 and Monitoring Reactor Coolant Pressure Boundary in Section 3.1.16, which contains provisions equivalent to 10 CFR [Part] 50, Appendix A, GDC 14 and 30. This difference has no impact on the applicability of TSTF-554, Revision 1, to the ONS TS.

The affected portions of the ONS TS have some editorial differences from the corresponding STS on which TSTF-554 was based. None of the differences affect the applicability of TSTF-554 to the ONS TS.

ONS TS 1.1 Definition for Identified LEAKAGE, Item a.3, already ends with a ; so the markups in this license amendment request differ editorially from TSTF-554 which shows the punctuation changing from a. to a ;.

ONS TS 1.1 Definition for Unidentified LEAKAGE Item b, ends with a. instead of a ; and does not end with and as shown in TSTF-554 and the STS. The ; and and has been added for consistency with the STS.

The licensee further states:

ONS TS 3.4.13, RCS Leakage, Required Action B.1 (new Required Action C.1), states the Unit must be in MODE 3 with a Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , whereas the Completion Time for the equivalent Required Action B.1 in STS and TSTF-554, Revision 1 is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . This difference in the default Completion Time does not impact the applicability of TSTF-554, Revision 1 to the ONS TS.

RNP The licensee states:

TSTF-554, Revision 1, and its corresponding Safety Evaluation discuss the applicable regulatory requirements and guidance, including the applicability of 10 CFR [Part] 50, Appendix A, General Design Criteria (GDC) 14 and 30. RNP Unit 2 was not licensed to the current 10 CFR 50, Appendix A, GDC. RNP UFSAR Chapter 3, "Design of Structures, Components, Equipment and Systems," states RNP was evaluated with respect to the proposed GDC contained in the Federal Register notice published July 11, 1967 (ML043310029). The UFSAR contains a discussion of the criteria as well as a summary of the criteria by groups.

The RNP UFSAR provides criteria for the Reactor Coolant Pressure Boundary in Section 3.1.2.9 and Monitoring Reactor Coolant Leakage in Section 3.1.2.16, which contains provisions equivalent to 10 CFR [Part]

50, Appendix A, GDC 14 and 30. This difference has no impact on the applicability of TSTF-554, Revision 1, to the RNP TS.

The affected portions of the RNP TS have some editorial differences from the corresponding STS on which TSTF-554 was based. None of the differences affect the applicability of TSTF-554 to the RNP TS.

RNP TS 1.1 Definition for Identified LEAKAGE, Item a.3, already ends with a ; so the markups in this license amendment requestion differ editorially from TSTF-554 which shows the punctuation changing from a. to a ;.

RNP TS 1.1 Definition for Unidentified LEAKAGE, Item b, does not end with and as shown in TSTF-554 and the STS. The and has been added for consistency with the STS.

The licensee further states:

RNP TS 3.4.13, RCS Operational Leakage, LCO d. describes the limit of primary to secondary leakage through any one steam generator as 75 gallons per day. The STS and TSTF-554, Revision 1 reflect a limit of 150 gallons per day in LCO 3.4.13.d. This difference in the RNP TS is more conservative than the STS and does not impact the applicability of TSTF-554, Revision 1, to the RNP TS.

Technical Evaluation of the licensees variations described above:

The NRC staffs review finds the variations in the proposed changes and do not adversely affect, a) the LEAKAGE Definitions, b) the Reactor Coolant System Operational Leakage requirements as specified in the Duke Fleets TSs, and c) the staffs technical evaluation provided in 3.1 of this SE for the applicability of TSTF-554 to the Fleets TSs, and therefore the variations are acceptable.

3.3 TS Change Consistency The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing guidance for customary terminology and formatting in Chapter 16.0 of the SRP and NUREG-1431. The NRC staff finds that the proposed changes are consistent with these documents and are therefore acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, States of North Carolina and South Carolina officials were notified of the proposed issuance of the amendments on July 12, 2023. On July 12 and 19, 2023, respectively, the State officials confirmed they had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement 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. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (88 FR 38547 dated June 13, 2023). Accordingly, 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 amendment(s) will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Ravi Grover, NRR/DSS Date: August 29, 2023

ML23195A078 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DSS/STSB/BC NAME SWilliams RButler VCusumano DATE 7/12/2023 7/17/2023 7/7/2023 OFFICE OGC - NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME DRoth DWrona SWilliams DATE 7/27/2023 8/29/2023 8/29/2023

v t e Letter Sequence Approval
EPID:L-2023-LLA-0028, Technical Specifications Task Force (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval, Approval Other:
MONTHYEAR2023-02-16 [Table View]

ML23195A078

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1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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