ML25134A248
| ML25134A248 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah, Watts Bar  |
| Issue date: | 12/18/2025 |
| From: | Kimberly Green Plant Licensing Branch II |
| To: | Erb D Tennessee Valley Authority |
| Buckberg, P | |
| References | |
| EPID L-2024-LLA-0152 | |
| Download: ML25134A248 (0) | |
Text
December 18, 2025 Mr. Delson C. Erb Vice President, OPS Support Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2; WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 374, 369, 178, AND 83 REGARDING TECHNICAL SPECIFICATION LIMITING CONDITION FOR OPERATION 3.5.2, ECCS - OPERATING, NOTE 1 TO INCLUDE RESIDUAL HEAT REMOVAL PUMP FLOW PATHS (EPID L-2024-LLA-0152)
Dear Mr. Erb:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 374 to Renewed Facility Operating License No. DPR-77, and Amendment No. 369 to Renewed Facility Operating License No. DPR-79, for the Sequoyah Nuclear Plant (SQN), Units 1 and 2, respectively; and Amendment No. 178 to Facility Operating License No.
NPF-90, and Amendment No. 83 to Facility Operating License No. NPF-96, for the Watts Bar Nuclear Plant (WBN), Units 1 and 2, respectively. These amendments are in response to your application dated November 12, 2024, as supplemented by letter dated April 22, 2025.
The amendments revise SQN, Units 1 and 2, and WBN, Units 1 and 2, Technical Specification Limiting Condition for Operation 3.5.2, ECCS - Operating, Note 1 to add one residual heat removal pump flow path to the emergency core cooling system flow paths allowed to be isolated to perform pressure isolation valve testing.
D. Erb A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Kimberly J. Green, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-327, 50-328, 50-390, and 50-391
Enclosures:
- 1. Amendment No. 374 to DPR-77
- 2. Amendment No. 369 to DPR-79
- 3. Amendment No. 178 to NPF-90
- 4. Amendment No. 83 to NPF-96
- 5. Safety Evaluation cc: Listserv TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 374 Renewed License No. DPR-77 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated November 12, 2024, as supplemented by letter dated April 22, 2025, 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 Title 10 of the Code of Federal Regulations (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-77 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 374 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David 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: December 18, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.12.18 09:34:55 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 374 SEQUOYAH NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace page 3 of the Renewed Facility Operating License with the attached page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following pages of Appendix A, Technical Specifications, with the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Insert 3.5.2-1 3.5.2-1 3.5.2-2 3.5.2-2 3.5.2-3 3.5.2-3 (3)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (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; and (5)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the Sequoyah and Watts Bar Unit 1 Nuclear Plants.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The Tennessee Valley Authority is authorized to operate the facility at reactor core power levels not in excess of 3455 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 374 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Initial Test Program The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authoritys Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
a.
Elimination of any test identified in Section 14 of TVAs Final Safety Analysis Report as amended as being essential; b.
Modification of test objectives, methods, or acceptance criteria for any test identified in Section 14 of TVAs Final Safety Analysis Report as amended as being essential; Amendment No. 374 Renewed License No. DPR-77
ECCS - Operating 3.5.2 SEQUOYAH - UNIT 1 3.5.2-1 Amendment 334, 358, 374 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.
NOTES-------------------------------------------
1.
In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation valve testing per SR 3.4.14.1.
2.
In MODE 3, ECCS pumps may be made incapable of injecting to support transition into or from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or until the temperature of all RCS cold legs exceeds Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR plus 25°F, whichever comes first.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable.
A.1 Restore train(s) to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR In accordance with the Risk Informed Completion Time Program B. Required Action and associated Completion Time not met.
B.1 Be in MODE 3.
AND B.2 Be in MODE 4.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours
ECCS - Operating 3.5.2 SEQUOYAH - UNIT 1 3.5.2-2 Amendment 334, 374 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.
C.1 Enter LCO 3.0.3.
Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed.
Number Position Function FCV-63-1 Open RHR Suction from RWST FCV-63-22 Open SIS Discharge to Common Piping In accordance with the Surveillance Frequency Control Program SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.3 Verify ECCS piping is full of water.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.
In accordance with the Inservice Testing Program SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
In accordance with the Surveillance Frequency Control Program
ECCS - Operating 3.5.2 SEQUOYAH - UNIT 1 3.5.2-3 Amendment 334, 366, 374 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each mechanical stop is in the correct position.
Charging Pump Injection Throttle Valves Safety Injection Cold Leg Throttle Valves Safety Injection Hot Leg Throttle Valves63-582 63-550 63-542 63-583 63-552 63-544 63-584 63-554 63-546 63-585 63-556 63-548 In accordance with the Surveillance Frequency Control Program TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 369 Renewed License No. DPR-79 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated November 12, 2024, as supplemented by letter dated April 22, 2025, 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-79 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 369 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David 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: December 18, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.12.18 09:35:33 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 369 SEQUOYAH NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace page 3 of the Renewed Facility Operating License with the attached page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following pages of Appendix A, Technical Specifications, with the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Insert 3.5.2-1 3.5.2-1 3.5.2-2 3.5.2-2 3.5.2-3 3.5.2-3 (3)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (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; and (5)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the Sequoyah and Watts Bar Unit 1 Nuclear Plants.
C.
This renewed 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 Tennessee Valley Authority is authorized to operate the facility at reactor core power levels not in excess of 3455 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 369 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Initial Test Program The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authority's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
a.
Elimination of any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential; Amendment No. 369 Renewed License No. DPR-79
ECCS - Operating 3.5.2 SEQUOYAH - UNIT 2 3.5.2-1 Amendment 327, 352, 369 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.
NOTES-------------------------------------------
1.
In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation valve testing per SR 3.4.14.1.
2.
In MODE 3, ECCS pumps may be made incapable of injecting to support transition into or from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or until the temperature of all RCS cold legs exceeds Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR plus 25°F, whichever comes first.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable.
A.1 Restore train(s) to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR In accordance with the Risk Informed Completion Time Program B. Required Action and associated Completion Time not met.
B.1 Be in MODE 3.
AND B.2 Be in MODE 4.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours
ECCS - Operating 3.5.2 SEQUOYAH - UNIT 2 3.5.2-2 Amendment 327, 369 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.
C.1 Enter LCO 3.0.3.
Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed.
Number Position Function FCV-63-1 Open RHR Suction from RWST FCV-63-22 Open SIS Discharge to Common Piping In accordance with the Surveillance Frequency Control Program SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.3 Verify ECCS piping is full of water.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.
In accordance with the Inservice Testing Program SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
In accordance with the Surveillance Frequency Control Program
ECCS - Operating 3.5.2 SEQUOYAH - UNIT 2 3.5.2-3 Amendment 327, 360, 369 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each mechanical stop is in the correct position.
Charging Pump Injection Throttle Valves Safety Injection Cold Leg Throttle Valves Safety Injection Hot Leg Throttle Valves63-582 63-550 63-542 63-583 63-552 63-544 63-584 63-554 63-546 63-585 63-556 63-548 In accordance with the Surveillance Frequency Control Program TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 178 License No. NPF-90 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated November 12, 2024, as supplemented by letter dated April 22, 2025, 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 Facility Operating License No. NPF-90 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 178 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA 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 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: December 18, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.12.18 09:36:11 -05'00'
ATTACHMENT TO AMENDMENT NO. 178 WATTS BAR NUCLEAR PLANT, UNIT 1 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace page 3 of Facility Operating License No. NPF-90 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following page of Appendix A, Technical Specifications, with the attached page.
The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Pages Insert Pages 3.5-4 3.5-4
Facility License No. NPF-90 Amendment No. 178 (4)
TVA, 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, instrument calibration, or other activity associated with radioactive apparatus or components; and (5)
TVA, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.
(1)
Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 178 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15)
Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational.
(4)
Vehicle Bomb Control Program (Section 13.6.9 of SSER 20)
During the period of the exemption granted in paragraph 2.D.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented. TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.
ECCS - Operating 3.5.2 Watts Bar-Unit 1 3.5-4 Amendment 55, 178 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.
NOTES------------------------------------------
1.
In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation valve testing per SR 3.4.14.1.
2.
In MODE 3, the safety injection pumps and charging pumps may be made incapable of injecting to support transition into or from the Applicability of the LCO 3.4.12, Cold Overpressure Mitigation System (COMS) for up to four hours or until the temperature of all the RCS cold legs exceeds 375F, whichever occurs first.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more trains inoperable.
AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.
A.1 Restore train(s) to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B.
Required Action and associated Completion Time not met.
B.1 Be in MODE 3.
AND B.2 Be in MODE 4.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 83 License No. NPF-96 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated November 12, 2024, as supplemented by letter dated April 22, 2025, 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 Facility Operating License No. NPF-96 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 83 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA 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 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: December 18, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.12.18 09:36:48 -05'00'
ATTACHMENT TO AMENDMENT NO. 83 WATTS BAR NUCLEAR PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. NPF-96 DOCKET NO. 50-391 Replace page 3 of Facility Operating License No. NPF-96 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following page of the Appendix A Technical Specifications with the attached page.
The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Pages Insert Pages 3.5-3 3.5-3 Unit 2 Facility Operating License No. NPF-96 Amendment No. 83 C.
The license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act, and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.
(1)
Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 83 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
TVA shall implement permanent modifications to prevent overtopping of the embankments of the Fort Loudon Dam due to the Probable Maximum Flood by June 30, 2018.
(4)
FULL SPECTRUM LOCA Methodology shall be implemented when the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1.
(5)
By December 31, 2019, the licensee shall report to the NRC that the actions to resolve the issues identified in Bulletin 2012-01, Design Vulnerability in Electrical Power System, have been implemented.
(6)
The licensee shall maintain in effect the provisions of the physical security plan, security personnel training and qualification plan, and safeguards contingency plan, and all amendments made pursuant to the authority of 10 CFR 50.90 and 50.54(p).
(7)
TVA shall fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
The TVA approved CSP was discussed in NUREG-0847, Supplement 28, as amended by changes approved in License Amendment No. 7.
(8)
TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report for the facility, as described in NUREG-0847, Supplement 29, subject to the following provision:
ECCS - Operating 3.5.2 Watts Bar - Unit 2 3.5-3 Amendment 83 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.
NOTES-------------------------------------------
- 1. In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation valve testing per SR 3.4.14.1.
- 2. In MODE 3, the safety injection pumps and charging pumps may be made incapable of injecting to support transition into or from the Applicability of the LCO 3.4.12, Cold Overpressure Mitigation System (COMS) for up to four hours or until the temperature of all the RCS cold legs exceeds the COMS arming temperature specified in the PTLR, whichever occurs first.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable.
AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.
A.1 Restore train(s) to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Required Action and associated Completion Time not met.
B.1 Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND B.2 Be in MODE 4.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 374 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 AMENDMENT NO. 369 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 AMENDMENT NO. 178 TO FACILITY OPERATING LICENSE NO. NPF-90 AMENDMENT NO. 83 TO FACILITY OPERATING LICENSE NO. NPF-96 SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 TENNESSEE VALLEY AUTHORITY DOCKET NOS. 50-327, 50-328, 50-390, AND 50-391
1.0 INTRODUCTION
By application dated November 12, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML24317A243), as supplemented by letter dated April 22, 2025 (ML25112A281), the Tennessee Valley Authority ( the licensee) requested changes to the technical specifications (TSs) for Sequoyah Nuclear Plant (SQN or Sequoyah), Units 1 and 2, and Watts Bar Nuclear Plant (WBN or Watts Bar), Units 1 and 2. The proposed changes would revise SQN, Units 1 and 2, and WBN, Units 1 and 2, TS Limiting Condition for Operation (LCO) 3.5.2, ECCS [Emergency Core Cooling System] - Operating, Note 1, to include one residual heat removal (RHR) pump flow path to perform pressure isolation valve (PIV) testing in addition to the existing allowances for safety injection (SI) pump flow path isolation.
The supplement dated April 22, 2025, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on January 21, 2025 (90 FR 7194).
2.0 REGULATORY EVALUATION
2.1
System Description
SQN and WBN each have ECCS that are used to protect the reactor core following design basis accidents. As stated in the license amendment request (LAR), the ECCS is made up of three subsystemscentrifugal charging (CC) (high head), SI (intermediate head), and RHR (low head). Each subsystem consists of two independent and redundant trains capable of providing 100 percent required flow.
TS Surveillance Requirement (SR) 3.4.14.1 requires the routine testing of reactor coolant system (RCS) PIVs to ensure that leakage through each PIV is within specified limits with RCS pressure greater than or equal to 2215 psig (pounds per square inch gauge) and less than or equal to 2255 psig. Currently, this surveillance is performed in Mode 4 at a lower RCS pressure, which requires the licensee to perform calculations to extrapolate the projected leak rate at full RCS pressure. As stated in the LAR, the licensee would like to be able to perform SR 3.4.14.1 in Mode 3 to allow for RCS conditions and resultant valve seating forces to more closely represent those associated with the pressure range in SR 3.4.14.1, but the testing flow path required is not currently allowed because both SI pumps and one of the RHR pump flow paths would be isolated.
2.2 Requested Changes The licensee proposed the following revision to the SQN, Units 1 and 2, and WBN, Units 1 and 2, TS LCO 3.5.2, Note 1 (addition of text noted by underlining):
In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation valve testing per SR 3.4.14.1.
2.3 Applicable Regulatory Requirements and Guidance Under Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, whenever a holder of a license wishes to amend the license, including TSs in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commissions regulations.
The Commissions regulatory requirements related to the content of TSs are set forth in 10 CFR 50.36, Technical Specifications, which require, in pertinent part, that the TSs include:
(1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls.
The regulation at 10 CFR 50.36(a)(1) states, that Each applicant for a license authorizing operation of a... utilization facility shall include in his application proposed technical specifications with the requirements of this section. A summary statement of the bases or reasons for such specifications... shall also be included in the application, but shall not become part of the technical specifications.
As stated in 10 CFR 50.36(b), each license authorizing operation of a production or utilization facility will include technical specifications. The TSs will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.
Under 10 CFR 50.36(c)(2), the technical specifications will include LCOs, which are the lowest functional capability or performance level of equipment required for safe operation of the facility.
When LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCOs can be met.
The regulation at 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, requires the ECCS of applicable plants to be designed such that ECCS cooling performance following postulated loss-of-coolant accidents (LOCAs) conforms to the criteria set forth in 10 CFR 50.46(b). In addition, ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated. Alternatively, an ECCS evaluation model may be developed in conformance with Appendix K, ECCS Evaluation Models, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities. The regulation at 10 CFR 50.46(b) specifies five criteria to which calculated ECCS cooling performance following postulated LOCAs must conform.
The regulations in 10 CFR 50.34(a)(3) and Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, establish the minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety. The NRC staff considered the following GDC as part of its review:
GDC 35, Emergency core cooling, states:
A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.
GDC 37, Testing of emergency core cooling, states:
The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.
According to section 3.1.1 of the Watts Bar Dual-Unit Updated Final Safety Analysis Report (UFSAR) (ML23346A225), the plant was originally designed to comply with the Proposed General Design Criteria for Nuclear Power Plant Construction Permits, published in July 1967.
However, the licensee assessed the standards applied to the design, the plant design features, and plant procedures compared to the NRC GDC published as Appendix A to 10 CFR Part 50 in July 1971, as discussed in UFSAR section 3.1.2.
Similarly, per section 3.1.2 of the Sequoyah UFSAR (ML23349A014), the plant was also originally designed to comply with the intent of the July 1967 Proposed General Design Criteria for Nuclear Power Plant Construction Permits. However, the licensee addresses the NRC General Design Criteria published as Appendix A to 10 CFR 50 in July 1971, as outlined in UFSAR section 3.1.2. The two Appendix A GDC relevant to this LAR are listed above.
The NRC staffs guidance for the review of TSs is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]
Edition, chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared Standard Technical Specifications for each of the LWR nuclear designs. The SQN and WBN units are Westinghouse four-loop design reactors. Accordingly, the NRC staffs review included consideration of whether the proposed changes are consistent with NUREG-14311, as modified by NRC-approved travelers.
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensees application to determine whether the proposed changes are consistent with the regulations, guidance, and plant-specific design and licensing basis information discussed in section 2.3 of this safety evaluation (SE).
The LAR cites the credited function of the ECCS to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:
a) A pipe break or spurious valve lifting in the RCS which cause a discharge larger than that which can be made up by the normal makeup system, up to and including the instantaneous circumferential rupture of the largest pipe in the RCS (i.e., LOCA).
b) Rupture of a control rod drive mechanism causing a rod cluster control assembly ejection accident.
c)
A pipe break or spurious valve lifting in the secondary system, up to and including the instantaneous circumferential rupture of the largest pipe in the secondary system.
d) A steam generator tube rupture (SGTR).
These accidents are discussed further in each plants UFSAR, chapter 15.4, and are dispositioned individually below.
3.1 Evaluation of Potential Impact on LOCA The LAR proposes to add one RHR pump flow path (i.e., the flow path between both RHR pumps and two RCS cold legs, as indicated in Figure 1 of the LAR) to the flow paths authorized by LCO 3.5.2, Note 1, to be temporarily isolated to perform PIV testing per SR 3.4.14.1. Both SI pump flow paths are already authorized to be simultaneously isolated for this purpose by the same note.
1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21259A155 and ML21259A159, respectively).
The LOCA analyses for all four units, as discussed in section 15.4.1.1.4 in the WBN UFSAR and section 15.4.1 in the SQN UFSAR, assume that 100 percent of the ECCS flow equivalent to a single operable ECCS train is available upon accident initiation (i.e., one CC subsystem, one SI subsystem, and one RHR subsystem are each operable and capable of injecting into all four cold legs). The current WBN and SQN, TS LCO 3.5.2, Note 1 allows SI subsystems to be inoperable for a short duration for the purposes of conducting surveillance testing. Isolating one RHR pump flow path in addition to both SI pump flow paths, as requested in the LAR, also renders both RHR subsystems inoperable due to the inability of either RHR pump to inject into all four cold legs automatically and further reduces available ECCS below the assumptions associated with the existing LOCA analyses in the UFSARs. However, existing TS ECCS operability requirements were derived assuming event initiation from bounding full-power conditions, rather than hot standby (Mode 3) conditions, which are less limiting.
The licensee cites several shutdown scoping studies not formally approved by the NRC as the bases for stating that a large-break LOCA (break diameter exceeding 6 inches) in Mode 3 is not a credible accident. These studies also provide guidance for deriving allowable operator action times to restore the isolated ECCS flow paths in the event of a small-or medium-break LOCA in Mode 3 that occurs while SR 3.4.14.1 testing is being conducted under the allowances of the proposed LCO 3.5.2, Note 1. These scoping studies are cited from a Westinghouse notification letter discussing ECCS performance considerations for LOCA in Modes 3 and 4 (ML010920123); WCAP-12476, Revision 0, Evaluation of LOCA During Mode 3 and 4 Operation for Westinghouse NSSS (Nuclear Steam Supply System), a Westinghouse Owners Group topical report that was submitted for NRC review but later withdrawn, since the associated issues related to shutdown risk and shutdown operations are addressed by licensees through voluntary programs (see ML003670146 and ML003726027); and PWROG-19021, Evaluation of LOCA During Mode 3 and Mode 4 Operation for Westinghouse NSSS, which replaced Revision 1 of WCAP-12476 and was neither endorsed nor approved by the NRC. In summarizing the relevant conclusions of the most recent shutdown LOCA study in PWROG-19021, the LAR states that the initiation of a safety injection within 10 minutes of indication of the LOCA will preclude any meaningful reactor core uncovery, and, as a result, no significant core heat up will occur and the 10 CFR 50.46 criteria will be met.
Section 3.0 of the LAR states that sufficient ECCS capacity will continue to exist such that the 10 CFR 50.46 criteria will be met, regardless of the Mode 3 conditions from which a LOCA initiates. The LAR indicates that the ECCS accumulators will be automatically available at RCS pressures exceeding 1000 psi, both CC pumps will be automatically available to inject into all four cold legs, and one RHR pump will be automatically available to inject into two cold legs.
The NRC staff reviewed the Westinghouse notification letter discussed above (ML010920123),
as well as the assumptions made in section 3.0 of the LAR regarding available equipment and operator actions. The staff performed this review to determine whether the conclusions derived in the LAR, and the assumptions, upon which those conclusions were based, were adequately justified. Although the shutdown studies have not been formally approved by the NRC, the technical content contained within can provide information to inform licensee risk mitigation actions.
One of the primary conclusions stated in the Westinghouse letter is that Mode 3 and Mode 4 conditions (the plant statuses evaluated as part of the notification) have lower power, temperature, and pressure initial conditions than assumed in the analyses of record, which are evaluated at power. Therefore, the NRC staff finds that the licensees assumption that a large-break LOCA need not be considered in the analysis supporting the proposed change to LCO 3.5.2, Note 1, is supported.
The other primary conclusion derived in the Westinghouse letter is that, for credible Mode 3 and Mode 4 LOCAs, operator actions can be taken to avoid exceeding the 10 CFR 50.46 acceptance criteria. Certain combinations of those operator actions and ECCS equipment availability are assumed when concluding that ECCS performance would be acceptable during a credible Mode 3 or Mode 4 LOCA.
In the supplement dated April 22, 2025, the licensee confirmed that procedural guidance will exist for both SQN and WBN such that operators can initiate the necessary ECCS flow within the assumed timeframes described in the LAR and the cited shutdown LOCA studies, including when initial conditions are representative of those that will be present when SR 3.4.14.1 testing is being undertaken per the proposed Note 1 of LCO 3.5.2 (i.e., in Mode 3 with both SI flow paths and one RHR flow path isolated, and testing in progress). Thus, the NRC staff finds that reasonable assurance exists that the operators can perform the actions required within the timeframes necessary to meet the 10 CFR 50.46 acceptance criteria in the event of a small-break LOCA in Mode 3.
Operability of ECCS accumulators is addressed under LCO 3.5.1, Accumulators, rather than LCO 3.5.2, ECCS - Operating; thus, the proposed changes to Note 1 of LCO 3.5.2 (Attachments 1 and 2 of the LAR) will not ensure that ECCS accumulators are operable, when required by the Applicability of LCO 3.5.1 (i.e., for the purposes of the LAR, in Mode 3 with pressurizer pressure > 1000 psig), while invoking the provisions of that Note. Similarly, the proposed changes to Note 1 of LCO 3.5.2 (Attachments 1 and 2 of the licensee submittal) do not involve the CC subsystems; as such, no assurance is provided that the CC subsystems will remain operable for the 2-hour period during which the SI and RHR subsystems are inoperable.
However, in the supplement dated April 22, 2025, the licensee confirmed that procedural guidance for both SQN and WBN will exist to verify operability/availability of the ECCS subsystems that are assumed to be available and are otherwise unaffected by the proposed note (i.e., CC subsystems, ECCS accumulators > 1000 psig) prior to the initiation of, and for the duration of, SR 3.4.14.1 PIV testing undertaken in accordance with the proposed Note 1.
Consequently, the NRC staff finds that sufficient emergency core cooling will remain capable of being provided, even when testing under the proposed Note 1 of LCO 3.5.2 is underway.
As described in sections 6.3 and 15.4.1 of the SQN and WBN UFSARs, following the injection phase of a LOCA, the ECCS is also utilized in the recirculation mode to provide long-term core cooling. Manual actions are required to establish cold leg recirculation (and, later in the event progression, hot leg recirculation) such that, given the feasibility and timing of the operator actions and ongoing availability of the ECCS subsystems discussed above, the ECCS alignment associated with the proposed change to LCO 3.5.2 will not challenge the ability of the operators to establish recirculation when and as required.
In addition to its ECCS functions (i.e., operation in both the injection and recirculation modes of ECCS), the RHR system performs a containment heat removal function. As described in each plants UFSAR section 6.2.2, each unit has two containment spray subsystems, with each subsystem comprised of a containment spray train and an RHR spray train. The containment spray subsystem is necessary to spray cool water into the containment atmosphere in the event of a LOCA to ensure that containment pressure does not exceed its design limit. The LAR states that the containment spray function of RHR will not be affected by the proposed change to LCO 3.5.2, Note 1.
The NRC staff reviewed the design basis information for the containment spray RHR function presented in the WBN and SQN UFSARs, including system drawings and timing requirements for initiation of RHR spray following a LOCA. Based on this assessment, the staff determined that the containment spray function of RHR will remain unaffected by the changes presented in the LAR.
The purpose of the TSs is to assure that the plant is operated within the bounds of its analyses or, taken out of the condition to which the analyses apply in a reasonable time. The LAR states that large-break LOCAs do not need to be considered when evaluating the proposed change.
However, 10 CFR 50.46 requires that ECCS cooling performance be calculated for a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated. The regulation was written to assure that the most severe plant conditions were evaluated. It did not address operating conditions that are much less challenging than the typical Mode 1 conditions that are the most severe for LOCAs. Typically, the most severe postulated LOCA is a double-ended rupture of the largest pipe in the reactor coolant system, which is postulated to occur at full power (i.e.,
Mode 1). For this reason, most licensees do not have plant-specific Mode 3 LOCA analysis. The licensee provided significant justification, including precedents for similar changes at other plants, and industry studies of plant responses to LOCAs in Mode 3.
The licensees procedures (as reflected in the TS bases markups and confirmed in the supplement dated April 22, 2025) ensure that the isolated systems can be placed back into service from the control room. The proposed change would allow the plant to be operated for a limited time with one train of RHR isolated from the RCS in Mode 3. The NRC staff determined that operation in this condition for the allowed time will not result in a significant risk to the plant and will allow the plant to perform TS required testing under conditions that are as close to design as practical. Therefore, the NRC staff finds that, with the proposed change to TS LCO 3.5.2, GDC 37 will continue to be met.
3.2 Evaluation of Potential Impact on Rod Cluster Control Assembly (RCCA) Ejection The NRC staff reviewed the existing RCCA ejection analyses in section 15.4.6 of the WBN and SQN UFSARs to determine the nature and magnitude of any potential impact from the proposed LCO 3.5.2 change. The UFSAR analyses of RCCA ejections evaluated both full-power and zero-power cases, which analyzes the full range of RCCA ejections, so initiation of the event from Mode 3 conditions is already bounded. Based on event progression, availability and timing of SI injection have no effect on the initiation or immediate reactivity effects of an RCCA ejection.
Ejection of an RCCA causes a break in the reactor pressure vessel head, and thus represents a type of small-break LOCA. The UFSAR analyses and plant and operator responses to a LOCA and to subsequent RCS depressurization were previously evaluated in section 3.1 of this SE.
3.3 Evaluation of Potential Impact on Secondary System Pipe Break The NRC staff reviewed the existing analyses of secondary system pipe breaks in the WBN and SQN UFSARs to determine the nature and magnitude of any potential impact from the proposed LCO 3.5.2 change. The WBN and SQN UFSARs each evaluate the major rupture of a main steam line and the major rupture of a main feedwater pipe as part of assessing major secondary system pipe ruptures as limiting faults.
Per section 15.4.2.1.2 of each UFSAR, the RHR subsystems within the ECCS are not modeled and, thus, are neither relied upon nor assumed to function within the analysis of a major rupture of a main steam line. Consequently, the NRC staff determined that adding the temporary isolation of one RHR flow path to the allowances of LCO 3.5.2, Note 1, does not have any bearing on major main steam line ruptures.
Per section 15.4.2.2 of each UFSAR, a major main feedwater line rupture could either lead to an RCS heatup or an RCS cooldown, depending upon break size and plant conditions at the time of the rupture. RCS cooldown effects are bounded by the analysis of major main steam line ruptures. Major main feedwater line ruptures leading to RCS heatup are mitigated by the auxiliary feedwater system, rather than the ECCS. Therefore, the NRC staff determined that the proposed change to LCO 3.5.2, Note 1, also has no bearing on major feedwater line ruptures.
3.4 Evaluation of Potential Impact on SGTR The NRC staff reviewed the existing SGTR analyses in the WBN and SQN UFSARs to determine the nature and magnitude of any potential impact from the proposed LCO 3.5.2 change. Section 15.4.3 of the SQN and WBN UFSARs evaluates plant and operator responses to an SGTR. An SGTR that initiates from full power (rather than Mode 3) and is analyzed assuming conservative initial conditions and assumptions has been previously established to represent a reasonably conservative basis for an SGTR accident such that separate consideration of a zero power SGTR is not merited (Attachment 1 to WCAP-10698-P-A, not publicly available). During a bounding SGTR accident at SQN or WBN, the RHR pumps would start upon receipt of an SI signal but would not inject since RCS conditions are above their shutoff head. RHR would not be utilized during the SGTR until its decay heat removal function (as described in UFSAR section 5.5.7) is needed to bring the reactor into cold shutdown.
Therefore, the staff determined that the proposed change to LCO 3.5.2, Note 1, also has no bearing on SGTR accidents.
3.5 Technical Conclusion Based on its evaluation above, the NRC staff finds that the proposed revision to TS LCO 3.5.2, Note 1, will not affect the requirements of the LCO such that safety is adversely affected. If the licensee cannot exit the conditions allowed by the proposed Note 1 in LCO 3.5.2 in the required time, then it will have to comply with the appropriate Required Actions of TS 3.5.2.
Additionally, the NRC staff determined that most limiting LOCA will not be affected by the proposed TS change and, therefore, the licensee will continue to comply with 10 CFR 50.46.
The staff also determined that the proposed change to Note 1 does not involve any changes to the ECCS design; therefore, GDC 35 will continue to be met. Lastly, the staff determined that, with the proposed change to TS LCO 3.5.2, GDC 37 will continue to be met.
Therefore, the NRC staff finds that the proposed change to TS LCO 3.5.2, Note 1, is acceptable because the TS, as revised, will continue to meet the requirements in 10 CFR 50.36(c)(2).
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendment on May 13, 2025. The State official 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 previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on January 21, 2025 (90 FR 7194), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: J. Ambrosini, NRR C. Szumski, NRR S. Smith, NRR Date: December 18, 2025
ML25134A248 NRR-058 OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DSS/SNSB/BC NAME KGreen CAdams (SL)
DMurdock DATE 5/14/2025 5/20/2025 5/12/2025 OFFICE NRR/DSS/STSB/BC NRR/DORL/LPLII-2/BC NRR/DORL/LPLII-2/PM NAME SMehta DWrona KGreen DATE 5/29/2025 12/18/2025 12/18/2025