ML22348A005
| ML22348A005 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, Watts Bar, Sequoyah  |
| Issue date: | 01/25/2023 |
| From: | Perry Buckberg Plant Licensing Branch II |
| To: | Jim Barstow Tennessee Valley Authority |
| Green K | |
| References | |
| EPID L-2022-LLA-0100 | |
| Download: ML22348A005 (54) | |
Text
January 25, 2023 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3; SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2; WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 327, 350, AND 310; 362 AND 356; 159 AND 67 REGARDING ADOPTION OF TSTF-554, REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS (EPID L-2022-LLA-0100)
Dear Mr. Barstow:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 327, 350, and 310 to Renewed Facility Operating Licenses Nos. DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant, Units 1, 2, and 3, respectively; Amendment Nos. 362 and 356 to Renewed Facility Operating License Nos. DPR-77 and DPR-79, for the Sequoyah Nuclear Plant, Units 1 and 2, respectively; and Amendment Nos. 159 and 67 to Facility Operating License Nos. NPF-90 and NPF-96 for the Watts Bar Nuclear Plant, Units 1 and 2, respectively. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 13, 2022.
The amendments revise the TSs to adopt Technical Specifications Task Force (TSTF) traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements, dated January 16, 2020, using the Consolidated Line Item Improvement Process (CLIIP). The NRC issued a Final Safety Evaluation approving TSTF-554 on December 18, 2020.
A copy of the Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Perry H. Buckberg, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260, 50-296, 50-327, 50-328, 50-390, and 50-391
Enclosures:
- 1. Amendment No. 327 to DPR-33
- 2. Amendment No. 350 to DPR-52
- 3. Amendment No. 310 to DPR-68
- 4. Amendment No. 362 to DPR-77
- 5. Amendment No. 356 to DPR-79
- 6. Amendment No. 159 to NPF-90
- 7. Amendment No. 67 to NPF-96
- 8. Safety Evaluation cc: Listserv
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 327 Renewed License No. DPR-33
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (the licensee) dated July 13, 2022, 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 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-33 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 327, are hereby incorporated in the renewed operating 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 30 days from the date 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: January 25, 2023 David J.
Wrona Digitally signed by David J. Wrona Date: 2023.01.25 07:41:24 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 327 BROWNS FERRY NUCLEAR PLANT, UNIT 1 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Replace page 3 of Renewed Facility Operating License No. DPR-33 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 the Appendix A, Technical Specifications, with the attached 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 3.4-9 3.4-9 3.4-10 3.4-10 BFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 327 (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 equipment and instrument calibration or associated with radioactive apparatus or components; (5)
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 the operation of the facility.
C.
This renewed operating 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; 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 steady state reactor core power levels not in excess of 3952 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 327, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 234 to Facility Operating License DPR-33, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 234. For SRs that existed prior to Amendment 234, 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 234.
Definitions 1.1 BFN-UNIT 1 1.1-4 (continued)
Amendment No. 234, 327 1.1 Definitions (continued)
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; b.
Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; c.
Total LEAKAGE Sum of the identified and unidentified LEAKAGE; 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 GENERATION RATE (LHGR)
The LHGR shall be the heat generation rate per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.
RCS Operational LEAKAGE 3.4.4 BFN-UNIT 1 3.4-9 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. d 5 gpm unidentified LEAKAGE; and
- c. d 30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and
- d. d 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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
deactivated automatic valve, blind flange, or check valve.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. Unidentified LEAKAGE not within limit.
OR Total LEAKAGE not within limit.
B.1 Reduce LEAKAGE to within limits.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (continued)
Amendment No. 234, 327
RCS Operational LEAKAGE 3.4.4 BFN-UNIT 1 3.4-10 Amendment No. 234, 327 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Unidentified LEAKAGE increase not within limit.
C.1 Reduce LEAKAGE increase to within limits.
OR 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C.2 Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. Required Action and associated Completion Time not met.
D.1 Be in MODE 3.
AND D.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 /> 36 hours
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 350 Renewed License No. DPR-52
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (the licensee) dated July 13, 2022, 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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the 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-52 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 350, are hereby incorporated in the renewed operating 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 30 days from the date 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: January 25, 2023 David J.
Wrona Digitally signed by David J.
Wrona Date: 2023.01.25 07:41:50
-05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 350 BROWNS FERRY NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Replace page 3 of Renewed Facility Operating License No. DPR-52 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 the Appendix A, Technical Specifications, with the attached 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 3.4-9 3.4-9 3.4-10 3.4-10 BFN-UNIT 2 Renewed License No. DPR-52 Amendment No. 350 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 equipment and instrument calibration or associated with radioactive apparatus or components; (5)
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 the operation of the facility.
C.
This renewed operating 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; 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 steady state reactor core power levels not in excess of 3952 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 350, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 253 to Facility Operating License DPR-52, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 253. For SRs that existed prior to Amendment 253, 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 253.
- 3)
The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's
Definitions 1.1 BFN-UNIT 2 1.1-4 (continued)
Amendment No. 253, 350 1.1 Definitions (continued)
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; b.
Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; c.
Total LEAKAGE Sum of the identified and unidentified LEAKAGE; 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 GENERATION RATE (LHGR)
The LHGR shall be the heat generation rate per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.
RCS Operational LEAKAGE 3.4.4 BFN-UNIT 2 3.4-9 Amendment No. 253, 350 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.
d 5 gpm unidentified LEAKAGE; and c.
d 30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and d.
d 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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
deactivated automatic valve, blind flange, or check valve.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. Unidentified LEAKAGE not within limit.
OR Total LEAKAGE not within limit.
B.1 Reduce LEAKAGE to within limits.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (continued)
RCS Operational LEAKAGE 3.4.4 BFN-UNIT 2 3.4-10 Amendment No. 253, 350 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Unidentified LEAKAGE increase not within limit.
C.1 Reduce LEAKAGE increase to within limits.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> OR C.2 Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. Required Action and associated Completion Time not met.
D.1 Be in MODE 3.
AND D.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 /> 36 hours
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 310 Renewed License No. DPR-68
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (the licensee) dated July 13, 2022, 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 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-68 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 310, are hereby incorporated in the renewed operating 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 30 days from the date 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: January 25, 2023 David J.
Wrona Digitally signed by David J.
Wrona Date: 2023.01.25 07:42:19
-05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 310 BROWNS FERRY NUCLEAR PLANT, UNIT 3 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Replace page 3 of Renewed Facility Operating License No. DPR-68 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 the Appendix A, Technical Specifications, with the attached 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 3.4-9 3.4-9 3.4-10 3.4-10 BFN-UNIT 3 Renewed License No. DPR-68 Amendment No. 310 (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 equipment and instrument calibration or associated with radioactive apparatus or components; (5)
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 the operation of the facility.
C.
This renewed operating 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; 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 steady state reactor core power levels not in excess of 3952 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 310, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 212 to Facility Operating License DPR-68, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 212. For SRs that existed prior to Amendment 212, 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 212.
Definitions 1.1 BFN-UNIT 3 1.1-4 (continued)
Amendment No. 212, 310 1.1 Definitions (continued)
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; b.
Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; c.
Total LEAKAGE Sum of the identified and unidentified LEAKAGE; 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 GENERATION RATE (LHGR)
The LHGR shall be the heat generation rate per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.
RCS Operational LEAKAGE 3.4.4 BFN-UNIT 3 3.4-9 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. d 5 gpm unidentified LEAKAGE; and
- c. d 30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and
- d. d 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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
deactivated automatic valve, blind flange, or check valve.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. Unidentified LEAKAGE not within limit.
OR Total LEAKAGE not within limit.
B.1 Reduce LEAKAGE to within limits.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (continued)
Amendment No. 212, 310
RCS Operational LEAKAGE 3.4.4 BFN-UNIT 3 3.4-10 Amendment 1R 212, 310 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Unidentified LEAKAGE increase not within limit.
C.1 Reduce LEAKAGE increase to within limits.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> OR C.2 Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. Required Action and associated Completion Time not met.
D.1 Be in MODE 3.
AND D.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 /> 36 hours
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 362 Renewed License No. DPR-77
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (the licensee) dated July 13, 2022, 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 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. 362, are hereby incorporated in the renewed 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 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 25, 2023 David J.
Wrona Digitally signed by David J.
Wrona Date: 2023.01.25 07:42:44
-05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 362 SEQUOYAH NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace page 3 of Renewed Facility Operating License No. DPR 77 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 the Appendix A, Technical Specifications, with the attached 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 Amendment No. 362 Renewed License No. DPR-77 (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. 362 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;
Definitions 1.1 SEQUOYAH - UNIT 1 1.1-3 Amendment 334, 352, 362 1.1 Definitions ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its 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 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.
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 SEQUOYAH - UNIT 1 3.4.13-1 Amendment 334, 362 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. Required Action and C.1 Be in MODE 3.
AND C.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours associated Completion Time notmet.
OR Primary to secondary LEAKAGE not within limit.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 356 Renewed License No. DPR-79
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (the licensee) dated July 13, 2022, 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-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. 356, are hereby incorporated in the renewed 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 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 25, 2023 David J.
Wrona Digitally signed by David J.
Wrona Date: 2023.01.25 07:43:11
-05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 356 SEQUOYAH NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace page 3 of Renewed Facility Operating License No. DPR 79 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 the Appendix A Technical Specifications with the attached 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 Amendment No. 356 Renewed License No. DPR-79 (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. 356 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;
Definitions 1.1 SEQUOYAH - UNIT
1.1-3 Amendment 3, 3, 356 1.1 Definitions ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its 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 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.
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 SEQUOYAH - UNIT
3.4.13-1 Amendment 3, 356 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. Required Action and associated Completion Time notmet.
OR C.1 Be in MODE 3.
AND C.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours Primary to secondary LEAKAGE not within limit.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 159 License No. NPF-90
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (TVA) dated July 13, 2022, 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 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. 159 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 30 days.
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 Facility Operating License and Technical Specifications Date of Issuance: January 25, 2023 David J.
Wrona Digitally signed by David J.
Wrona Date: 2023.01.25 07:43:34
-05'00'
ATTACHMENT TO AMENDMENT NO. 159 WATTS BAR NUCLEAR PLANT, UNIT 1 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace page 3 of Renewed Facility Operating License No. NPF 90 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 the Appendix A, Technical Specifications, with the attached 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 1.1-4 1.1-4 3.4-30 3.4-30 Unit 1 Facility License No. NPF-90 Amendment No. 159 (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. 159 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.
Definitions 1.1 1.1 Definitions (continued)
(continued)
Watts Bar-Unit 1 1.1-3 Amendment 24, 141, 146, 159 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.
La The maximum allowable primary containment leakage rate, La, shall be.25% of primary containment air weight per day at the calculated peak containment pressure (Pa).
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
Definitions 1.1 1.1 Definitions (continued)
Watts Bar-Unit 1 1.1-4 Amendment 65, 82, 159 LEAKAGE 3.
(continued)
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; 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.
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).
PDMS The Power Distribution Monitoring System (PDMS) is a real-time three dimensional core monitoring system. The system utilizes existing core instrumentation data and an on-line neutronics code to provide surveillance of core thermal limits.
PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
RCS Operational LEAKAGE 3.4.13 Watts Bar-Unit 1 3.4-30 Amendment 38, 65, 159 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 67 License No. NPF-96
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (TVA) dated July 13, 2022, 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.
Unit 2 Facility License No. NPF-96 Amendment No. 67 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. 67 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 30 days.
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 Operating License and Technical Specifications Date of Issuance: January 25, 2023 David J.
Wrona Digitally signed by David J.
Wrona Date: 2023.01.25 07:44:02
-05'00'
ATTACHMENT TO AMENDMENT NO. 67 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 pages of the Appendix A, Technical Specifications, with the attached 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 3.4-28 3.4-28 Unit 2 Facility Operating License No. NPF-96 Amendment No. 67 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. 67 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:
Definitions 1.1 1.1 Definitions (continued)
(continued)
Watts Bar - Unit 2 1.1-4 Amendment 67 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; 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 Watts Bar-Unit
3.4-
Amendment 67 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 327 TO RENEWED FACILITY OPERATING LICENSE NO. DPR 33 AMENDMENT NO. 350 TO RENEWED FACILITY OPERATING LICENSE NO. DPR 52 AMENDMENT NO. 310 TO RENEWED FACILITY OPERATING LICENSE NO. DPR 68 AMENDMENT NO. 362 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 AMENDMENT NO. 356 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 AMENDMENT NO. 159 TO FACILITY OPERATING LICENSE NO. NPF-90 AMENDMENT NO. 67 TO FACILITY OPERATING LICENSE NO. NPF-96 BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 TENNESSEE VALLEY AUTHORITY DOCKET NOS. 50-259, 50-260, 50-296, 50-327, 50-328, 50-390, AND 50-391
1.0 INTRODUCTION
By application dated July 13, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22196A366), Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request (LAR) for amendments to the Technical Specifications (TSs) for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Sequoyah Nuclear Plant (SQN), Units 1 and 2, and Watts Bar Nuclear Plant (WBN), Units 1 and 2, (collectively, the TVA Operating Fleet).
In its application, the licensee 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), and the associated NRC staff safety evaluation (SE) of TSTF-554 (ML20322A024).
Per the application, TVA is not proposing any variations from the TS changes described in TSTF-554-A, Revision 1 or the applicable parts of the NRC staffs safety evaluation dated December 18, 2020.
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/drywell 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 TVA fleet TSs in order to adopt TSTF-554:
For each facility, the TS 1.1 identified LEAKAGE definition in sub-section a.2 would be revised to remove the exclusion of pressure boundary leakage from identified leakage by deleting the word either and the phrase not to be pressure boundary LEAKAGE.
The pressure boundary LEAKAGE definition in TS 1.1c for SQN and WBN units and TS 1.1d for BFN units, 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 RCS Operational LEAKAGE ACTIONS in TS 3.4.4 for BFN units and TS 3.4.13 for SQN and WBN units, would be revised to add a new Condition A to isolate the pressure boundary leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Existing TS 3.4.4 Condition C for BFN units and TS 3.4.13 Condition B for SQN and WBN units, would be revised to be applicable should any Action of TS 3.4.4 (BFN) and TS 3.4.13 (SQN, WBN) not be met by deleting of Condition A or B. These Conditions would further be revised to delete the condition for when pressure boundary leakage exists because pressure boundary leakage would be addressed by the new Condition A.
Existing Conditions A, B, and C would be renumbered to reflect the new Condition A. In addition, the required actions associated with these conditions would be renumbered accordingly.
1.3 Additional Proposed TS Changes Per the application, the TVA is not proposing any variations from the TS changes described in TSTF-554-A, Revision 1 or the applicable parts of the NRC staffs safety evaluation dated December 18, 2020.
2.0 REGULATORY EVALUATION
The regulation at 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 when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs 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 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), dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared standard technical specifications (STSs) for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREGs 14311 and 1433,2 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 SQN and WBN units because these facilities are pressurized water reactors (PWRs), and BFN units are boiling-1 U.S. Nuclear Regulatory Commission Standard Technical Specifications, Westinghouse Plants, NUREG 1431, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ADAMS Accession Nos. ML12100A222 and ML12100A228, respectively),
2 U.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).
water reactors (BWRs), and the NRC staff approved the TSTF-554 changes for PWR and BWR designs. The NRC finds that the licensees proposed changes to section 1.1 of TSs for the subject facilities, 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, for RCS operational leakage (these LCOs address 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 10 CFR 50.2 and was consistent with RG 1.45. The NRC staff further found that proposed new Condition A on pressure boundary leakage including its associated Required Action and Completion Time, is 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 me.
Based on the discussions above, the NRC staff finds that the proposed changes to the TVA facilities TS 1.1 definition and LCOs 3.4.4 and 3.4.13 correctly specify the lowest functional capability or performance levels of equipment required for safe operation of the facility in accordance with 10 CFR 50.36(c)(2)(i). Also, the remedial actions to be taken until each LCO can be met provide protection to the health and safety of the public, thereby satisfying 10 CFR 50.36(c)(2)(i).
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 NUREGs-1431 and 1433. 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 Commission's regulations, the Tennessee State official and the Alabama State official were notified of the proposed issuance of the amendments on December 13, 2022, and January 5, 2023, respectively. The State officials 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, published in the Federal Register on September 6, 2022 (87 FR 54554), 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 amendment[s].
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: January 25, 2023
ML22348A005 OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DSS/STSB/BC NAME PBuckberg RButler VCusumano DATE 12/23/2022 12/23/2022 12/12/2022 OFFICE OGC - NLO NRR/DORL/LPLII-2/BC NRR/DORL/LPLII-2/PM NAME BAyersman DWrona PBuckberg DATE 01/06/2023 01/25/2023 01/25/2023