ML22276A161
| ML22276A161 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar, Sequoyah  |
| Issue date: | 10/24/2022 |
| From: | Kimberly Green Plant Licensing Branch II |
| To: | Jim Barstow Tennessee Valley Authority |
| Green K | |
| References | |
| EPID L-2022-LLA-0051 | |
| Download: ML22276A161 (42) | |
Text
October 24, 2022 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:
SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2; AND WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 359, 353, 155, AND 63 REGARDING ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE TRAVELER TSTF-577, REVISED FREQUENCIES FOR STEAM GENERATOR TUBE INSPECTIONS (EPID L-2022-LLA-0051)
Dear Mr. Barstow:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 359 and 353 to Renewed Facility Operating License Nos. DPR-77 and DPR-79, for the Sequoyah Nuclear Plant, Units 1 and 2, respectively; and Amendment Nos. 155 and 53 to Facility Operating License Nos. NPF-90 and NPF-96 for the Watts Bar Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated April 4, 2022, as supplemented by letter dated July 13, 2022.
The amendments revise the steam generator tube inspection frequencies and reporting requirements in the Steam Generator (SG) Program and the Steam Generator Tube Inspection Report technical specifications for Sequoyah Nuclear Plant and Watts Bar Nuclear Plant. These revisions are based on Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections.
A copy of the Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Kimberly J. Green, 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. 359 to DPR-77
- 2. Amendment No. 353 to DPR-79
- 3. Amendment No. 155 to NPF-90
- 4. Amendment No. 63 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. 359 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 April 4, 2022, as supplemented by letter 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. 359, 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 its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: October 24, 2022 David J.
Wrona Digitally signed by David J. Wrona Date: 2022.10.24 14:36:36 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 359 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 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 5.5-5 5.5-5 5.5-6 5.5-6 5.5-7 5.5-7 5.6-5 5.6-5
(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. 359 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. 359 Renewed License No. DPR-77
Programs and Manuals 5.5 SEQUOYAH - UNIT 1 5.5-5 Amendment 334,
5.5 Programs and Manuals 5.5.6 Inservice Testing Program (continued) d.
Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.
5.5.7 Steam Generator (SG) Program a.
Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.
Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.
b.
Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
1.
Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
AQ6* Program shallbe establishedand implementedto ensure that SG tube
integrity ismaintained. In addition,the6* Programshall include thefollowing:
Programs and Manuals 5.5 SEQUOYAH - UNIT 1 5.5-6 Amendment 334, 353,
5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) 2.
Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG.
3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
c.
Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2.
After the first refueling outage following SG installation, inspect 100%
of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.
Programs and Manuals 5.5 SEQUOYAH - UNIT 1 5.5-7 Amendment 334, 353,
5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) 3.
If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e.
Provisions for monitoring operational primary to secondary LEAKAGE.
Reporting Requirements 5.6 SEQUOYAH - UNIT 1 5.6-5 Amendment 334, 353, 356,
5.6 Reporting Requirements 5.6.5 Post Accident Monitoring Report When a report is required by Condition B or I of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.6 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, "Steam Generator (SG) Program." The report shall include:
D
The scope of inspections performed on each SG; E
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility; F
For each degradation mechanism found:
The nondestructive examination techniques utilized;
The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment;DQG
The number of tubes plugged during the inspection outage.
G
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results; H
The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;DQG I
The results of any SG secondary side inspections.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 353 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 April 4, 2022, as supplemented by letter 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. 353, 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 its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: October 24, 2022 David J.
Wrona Digitally signed by David J. Wrona Date: 2022.10.24 14:37:08 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 353 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 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 5.5-5 5.5-5 5.5-6 5.5-6 5.5-7 5.5-7 5.6-5 5.6-5 (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. 353 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. 353 Renewed License No. DPR-79
Programs and Manuals 5.5 SEQUOYAH - UNIT 2 5.5-5 Amendment 327,
5.5 Programs and Manuals 5.5.6 Inservice Testing Program (continued) d.
Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.
5.5.7 Steam Generator (SG) Program a.
Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.
Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.
b.
Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
1.
Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
AQ6* Program shallbe establishedand implementedto ensure that SG tube
integrity ismaintained. In addition,the6* Programshall include thefollowing:
Programs and Manuals 5.5 SEQUOYAH - UNIT 2 5.5-6 Amendment 327,
5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) 2.
Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG.
3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
c.
Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2.
After the first refueling outage following SG installation, inspect 100%
of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.
Programs and Manuals 5.5 SEQUOYAH - UNIT 2 5.5-7 Amendment 327,
5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) 3.
If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e.
Provisions for monitoring operational primary to secondary LEAKAGE.
Reporting Requirements 5.6 SEQUOYAH - UNIT 2 5.6-5 Amendment 327, 349,
5.6 Reporting Requirements 5.6.5 Post Accident Monitoring Report When a report is required by Condition B or I of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.6 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, "Steam Generator (SG) Program." The report shall include:
a.
The scope of inspections performed on each SG; b.
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility; c.
For each degradation mechanism found:
1.
The nondestructive examination techniques utilized; 2.
The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported; 3.
A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and 4.
The number of tubes plugged during the inspection outage.
d.
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results; e.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; and f.
The results of any SG secondary side inspections.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 155 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, the licensee) dated April 4, 2022, as supplemented by letter 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. 155 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.
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: October 24, 2022 David J.
Wrona Digitally signed by David J. Wrona Date: 2022.10.24 14:37:50 -04'00'
ATTACHMENT TO AMENDMENT NO. 155 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 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 5.0-15 5.0-15 5.0-16 5.0-16a 5.0-16 5.0-32 5.0-32
Amendment No. 155 Facility License No. NPF-90 (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. 155 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.
Procedures, Programs, and Manuals 5.7 (continued)
Watts Bar-Unit 1 5.0-15 Amendment 27, 38, 44, 65, 147,
5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Program An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:
a.
Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.
b.
Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
1.
Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
2.
Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage for all degradation mechanisms is not to exceed 150 gpd for each unfaulted SG. Leakage for all degradation mechanisms is not to exceed 1 gpm in the faulted SG.
Procedures, Programs, and Manuals 5.7 (continued)
Watts Bar-Unit 1 5.0-16 Amendment 27, 38, 44, 65, 147,
5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG) Program (continued) 3.
The operational leakage performance criterion is specified in LCO 3.4.13, RCS Operational LEAKAGE.
c.
Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2.
After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.
3.
If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e.
Provisions for monitoring operational primary-to-secondary LEAKAGE.
Record Retention 5.10 Watts Bar-Unit 1 5.0-32 Amendment 27, 38, 65, 96, 147,
5.9 Reporting Requirements (continued) 5.9.7 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days. Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.
5.9.8 PAMSRHSRUW 5.9.9 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.7.2.12,
Steam Generator (SG) Program. The report shall include:
D
The scope of inspections performed on each SG; E
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility; F
For each degradation mechanism found:
The nondestructive examination techniques utilized;
The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percentthrough-wall, only the total number of indications needs to be reported;
A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment;DQG
The number of tubes plugged during the inspection outage.
G
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results; H
The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;DQG I
The results of any SG secondary side inspections.
When a Report is required by Condition B or F of LCO 3.3.3, Post Accident Monitoring (PAM)
Instrumentation, a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 63 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, the licensee) dated April 4, 2022, as supplemented by letter 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-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. 63 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.
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: October 24, 2022 David J.
Wrona Digitally signed by David J. Wrona Date: 2022.10.24 14:38:18 -04'00'
ATTACHMENT TO AMENDMENT NO. 63 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 5.0-15 5.0-15 5.0-16 5.0-16 5.0-17 5.0-17a 5.0-17 5.0-35 5.0-35 5.0-36 5.0-36 Unit 2 Facility Operating License No. NPF-96 Amendment No. 63 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. 63 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:
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)
(continued)
Watts Bar - Unit 2 5.0-15 Amendment 40, 60,
5.7.2.12 Steam Generator (SG) Program An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:
D
Provisions for condition monitoring assessments. Condition
monitoring assessment means an evaluation of the "as found"
condition of the tubing with respect to the performance criteria for
structural integrity and accident induced leakage. The "as found"
condition refers to the condition of the tubing during a SG
inspection outage, as determined from the inservice inspection
results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted duringeach outage during which the SG tubes are inspected or plugged,to confirm that the performance criteria are being met.
E
Performance criteria for SG tube integrity. SG tube integrity shall
be maintained by meeting the performance criteria for tube
structural integrity, accident induced leakage, and operational
LEAKAGE.
Structural integrity performance criterion: All in-service SG
tubes shall retain structural integrity over the full range of
normal operating conditions (including startup, operation in
the power range, hot standby, and cooldown), all anticipated
transients included in the design specification and design
basis accidents. This includes retaining a safety factor of 3.0
against burst under normal steady state full power operation
primary-to-secondary pressure differential and a safety factor
of 1.4 against burst applied to the design basis accident
primary-to-secondary pressure differentials. Apart from the
above requirements, additional loading conditions associated
with the design basis accidents, or combination of accidents
in accordance with the design and licensing basis, shall also
be evaluated to determine if the associated loads contribute
significantly to burst or collapse. In the assessment of tube
integrity, those loads that do significantly affect burst or
collapse shall be determined and assessed in combination
with the loads due to pressure with a safety factor of 1.2 on
the combined primary loads and 1.0 on axial secondary
loads.
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)
Watts Bar - Unit 2 5.0-16 Amendment 2, 28,40, 60,
5.7.2.12 Steam Generator (SG) Program (continued) 2.
Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than an SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage for all degradation mechanisms is not to exceed 150 gpd for each unfaulted SG. Leakage for all degradation mechanisms is not to exceed 1 gpm in the faulted SG.
3.
The operational leakage performance criterion is specified in LCO 3.4.13, RCS Operational LEAKAGE.
c.
Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)
Watts Bar - Unit 2 5.0-17 Amendment 2, 28, 40, 60,
5.7.2.12 Steam Generator (SG) Program (continued) d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2.
After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.
- 3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e.
Provisions for monitoring operational primary-to-secondary LEAKAGE.
Reporting Requirements 5.9 5.9 Reporting Requirements (continued)
(continued)
Watts Bar-Unit 2 5.0-35 Amendment 40, 60,
5.9.7 DG Failures Report If an individual diesel generator (DG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that DG in that time period shall be reported within 30 days. Reports on DG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.
5.9.8 PAMS Report When a Report is required by Condition B or F of LCO 3.3.3, Post Accident Monitoring (PAM) Instrumentation, a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.9.9 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.7.2.12, Steam Generator (SG) Program. The report shall include:
D
The scope of inspections performed on each SG; E
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility; F
For each degradation mechanism found:
The nondestructive examination techniques utilized;
The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment;DQG
The number of tubes plugged during the inspection outage.
G
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;
Reporting Requirements 5.9 5.9 Reporting Requirements (continued)
Watts Bar-Unit 2 5.0-36 (continued)
Amendment 28, 40, 60,
5.9.9 Steam Generator Tube Inspection Report (continued)
H
The number and percentage of tubes plugged to date, and the effective
plugging percentage in each SG;DQG I
The results of any SG secondary side inspections
5.10 Record Retention (removed from Technical Specifications)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 359 TO RENEWED FACILITY OPERATING LICENSE DPR-77 AMENDMENT NO. 353 TO RENEWED FACILITY OPERATING LICENSE DPR-79 AMENDMENT NO. 155 TO FACILITY OPERATING LICENSE NPF-90 AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE 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 letter dated April 4, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22095A023), as supplemented by letter dated July 13, 2022 (ML22196A363), the Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC or Commission) for the Sequoyah Nuclear Plant (Sequoyah), Units 1 and 2; and Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2. The requested changes would revise the Steam Generator (SG) Program and the Steam Generator Tube Inspection Report technical specifications (TS) based on Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (ML21060B434), and the associated NRC staff safety evaluation (SE) of TSTF-577 (ML21098A188). The licensee requested that the NRC process the proposed amendments under the Consolidated Line Item Improvement Process (CLIIP).
The supplement dated July 13, 2022, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on June 14, 2022 (87 FR 36009).
The tubes within an SG function as an integral part of the reactor coolant pressure boundary and, in addition, isolate fission products in the primary coolant from the secondary coolant and the environment. Steam generator tube integrity means that the tubes are capable of performing this safety function in accordance with the plant design and licensing basis.
The Sequoyah and Watts Bar SGs have Alloy 690 thermally treated (Alloy 690TT) tubes.
2.0 REGULATORY EVALUATION
2.1 Requested Changes In accordance with the NRC staff-approved TSTF-577, the licensee proposed changes that would revise Sequoyah, Units 1 and 2, TS 5.5.7, Steam Generator (SG) Program, and TS 5.6.6, Steam Generator Tube Inspection Report, and Watts Bar, Units 1 and 2, TS 5.7.2.12, Steam Generator (SG) Program, and TS 5.9.9, Steam Generator Tube Inspection Report.
Specifically, the licensee proposed the following changes to adopt TSTF-577:
TS 5.5.7, Steam Generator (SG) Program (Sequoyah) and TS 5.7.2.12, Steam Generator (SG) Program (Watts Bar):
The introductory paragraph to TS 5.5.7 and TS 5.7.2.12 would be revised by replacing steam generator with SG in a couple instances.
TS 5.5.7.b.1 and TS 5.7.2.12.b.1 would be revised by replacing steam generator with SG in one instance.
TS 5.5.7.d.2 and TS 5.7.2.12.d.2 would be revised by deleting the requirement to base the inspection frequency on the more restrictive metric between either the effective full power months (EFPM) or refueling outage and to use just the EFPM metric (Sequoyah, Unit 2, and Watts Bar, Unit 2, only).
TS 5.5.7.d.2 and TS 5.7.2.12.d.2 would be revised by deleting the requirement to inspect 100 percent of the tubes during each period in paragraphs d.2.a and d.2.b (144 and 96 EFPM, respectively) and by adding the requirement to inspect 100 percent of the tubes every 96 EFPM (note, 96 EFPM currently permitted) (Sequoyah, Unit 1, and Watts Bar, Unit 1, only).
TS 5.5.7.d.2 and TS 5.7.2.12.d.2 would be revised by deleting the requirement to inspect 100 percent of the tubes during each period in paragraphs d.2.a, d.2.b, d.2.c, and d.2.d (144, 120, 96, and 72 EFPM, respectively) and by adding the requirement to inspect 100 percent of the tubes every 96 EFPM (Sequoyah, Unit 2, and Watts Bar, Unit 2, only).
TS 5.5.7.d.2 and TS 5.7.2.12.d.2 would be revised by deleting the allowance to extend the inspection period by 3 effective full power months and by deleting the discussion of prorating inspections.
TS 5.5.7.d.3 and TS 5.7.2.12.d.3 would be revised by replacing shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections) with shall be at the next refueling outage.
TS 5.6.6, Steam Generator Tube Inspection Report (Sequoyah) and TS 5.9.9, Steam Generator Tube Inspection Report (Watts Bar):
Existing reporting requirement b. would be renumbered as c. and be revised by editorial and punctuation changes.
New reporting requirement b. would be added to require the nondestructive examination (NDE) techniques utilized for tubes with increased degradation susceptibility be reported.
Existing reporting requirement c. would be renumbered as c.1. and be revised by editorial and punctuation changes.
Existing reporting requirement d. would be renumbered as c.2. and be revised to note that the location, orientation (if linear), measured size (if available), and voltage response do not need to be reported for tube wear indications at support structures that are less than 20 percent through-wall. However, the total number of tube wear indications at support structures that are less than 20 percent through-wall would be reported.
New reporting requirement d. would be added to require an analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection relative to the applicable performance criteria, including the analysis methodology, inputs, and results.
Existing reporting requirement e. would be renumbered as c.4. and be revised by editorial and punctuation changes.
Existing reporting requirement f. and h. would be combined, be renumbered as e.
(Sequoyah), and existing reporting requirement f. would be renumbered as e. (Watts Bar) and be revised by editorial and punctuation changes.
New reporting requirement f. would be added to require the results of any SG secondary side inspections be reported.
Existing reporting requirement g. would be renumbered as c.3. and be revised to add the requirements to report a description of the condition monitoring assessment, the margin to the tube integrity performance criteria, and a comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment. In addition, the requirement to report the results of tube pulls and in-situ testing would be deleted.
2.2 Additional Proposed TS Changes In addition to the changes proposed consistent with the traveler discussed in section 1.1, the licensee proposed the following variations.
2.2.1 Editorial Variations In license amendment request (LAR) section 2.2, Variations, the licensee identified editorial variations where the Sequoyah, Units 1 and 2, and Watts Bar, Units 1 and 2, TSs use different numbering than the standard technical specifications (STS) on which TSTF-577 was based.
Specifically, the Steam Generator (SG) Program is numbered 5.5.7 in the Sequoyah, Units 1 and 2, TSs, and is numbered 5.7.2.12 in the Watts Bar, Units 1 and 2, TSs, rather than 5.5.9 as stated in TSTF-577. In addition, the Steam Generator Tube Inspection Report, is numbered 5.6.6 in Sequoyah, Units 1 and 2, TSs, and is numbered 5.9.9 in the Watts Bar, Units 1 and 2, TSs, rather than 5.6.7 as stated in TSTF-577.
The NRC staff identified one additional editorial variation for the Sequoyah TSs. The licensee deleted the word provisions from Sequoyah, Units 1 and 2, TS 5.5.7 introductory paragraph.
In the LAR supplement dated July 13, 2022 (ML22196A363), the licensee proposed two additional editorial corrections to the Watts Bar Units 1 and 2, TSs. First, Watts Bar, Units 1 and 2, TS 5.9.9 is revised to add quotation marks for the title of Watts Bar, Units 1 and 2, TS 5.7.2.12. Second, Watts Bar, Unit 2, TS 5.7.2.12 paragraph b.1 is revised to add a comma after design specification to be consistent with the STS and the Watts Bar, Unit 1, TS.
2.2.2 Other Variations In LAR section 2.2, the licensee identified that Sequoyah, Unit 1, and Watts Bar, Units 1 and 2, TSs contain a few requirements that differ from the STS on which TSTF-577 was based.
Sequoyah, Unit 1, TS 5.5.7.d.2, and Watts Bar, Unit 1, TS 5.7.2.12.d.2, currently permit a permanent SG tube inspection frequency of every 96 effective full power months, which is consistent with TSTF-577, and is being retained.
Sequoyah, Unit 1, TS 5.5.7.d.2 and Watts Bar, Unit 1, TS 5.7.2.12.d.2, contain a requirement related to inspection probe technology and inspection techniques. This requirement is being deleted for consistency with TSTF-577.
Sequoyah, Unit 1, TS 5.6.6, and Watts Bar, Unit 1, TS 5.9.9, contain a requirement to discuss trending of tube degradation. This requirement is being deleted for consistency with TSTF-577 and because a similar requirement exists in the revised Sequoyah, Unit 1, TS 5.6.6.c, and Watts Bar, Unit 1, TS 5.9.9.c.
Westinghouse STS 5.5.9, Steam Generator (SG) Program, item b.2 contained in TSTF-577, Revision 1, states, Leakage is not to exceed [1 gpm] per SG. Whereas the Watts Bar, Units 1 and 2, TS 5.7.2.12.b.2 states, Leakage for all degradation mechanisms is not to exceed 150 gpd [gallons per day] for each unfaulted SG. Leakage for all degradation mechanisms is not to exceed 1 gpm in the faulted SG.
2.3 Regulatory Requirements and Guidance The regulations in Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.36(c)(5),
Administrative controls, state that [a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in
[10 CFR] 50.4. Technical Specification Section 5.0, Administrative Controls, requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained.
Programs established by the licensee, including the SG Program, are listed in the administrative controls section of the TS to operate the facility in a safe manner.
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 (SRP), 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 STSs for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREG-14311, as modified by NRC-approved travelers.
TSTF-577 revised the STSs related to SG tube inspections and SG tube inspection reporting requirements. The NRC approved TSTF-577, under the CLIIP on April 14, 2021 (ML21099A086).
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensees application to determine if the proposed changes are consistent with the regulations and guidance discussed in section 2.3 of this SE.
3.1 Proposed TS Changes to Adopt TSTF-577 The NRC staff compared the licensees proposed TS changes in section 2.1 of this SE against the changes approved in TSTF-577. In accordance with SRP Chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-577 are applicable because Sequoyah, Units 1 and 2, and Watts Bar, Units 1 and 2 are pressurized water reactor (PWR) design plants and the NRC staff approved the TSTF-577 changes for PWR designs. The NRC staff finds that the licensees proposed changes to the Sequoyah, Units 1 and 2, TSs, and Watts Bar, Units 1 and 2, TSs, in section 2.1 of this SE are consistent with those previously found acceptable in TSTF-577.
In the SE of TSTF-577, the NRC staff concluded that the TSTF-577 changes to STS 5.5.9, Steam Generator (SG) Program, and STS 5.6.7, Steam Generator Tube Inspection Report, were acceptable because, as discussed in section 3.0 of that SE, they continued to ensure SG tube integrity and, therefore, protected the public health and safety. In particular, the structural integrity performance criterion and accident-induced leakage performance criterion (explained in STS 5.5.9.b, items 1 and 2, respectively) will continue to be met with the proposed revised SG inspection intervals (maximum allowable time between SG inspections) and inspection periods (maximum allowable time between 100 percent of SG tubes inspections). Additionally, the proposed changes to the reporting requirements will provide more detailed and consistent information to the NRC. Therefore, the NRC staff found that the proposed changes to the SG program and inspection reporting requirements were acceptable because they continued to meet the requirements of 10 CFR 50.36(c)(5) by providing administrative controls necessary to assure operation of the facility in a safe manner. For these same reasons, the NRC staff concludes that the corresponding proposed changes to the Sequoyah and Watts Bar TSs in section 2.1 of this SE continue to meet the requirements of 10 CFR 50.36(c)(5).
3.2 Additional Proposed TS Changes 3.2.1 Editorial Variations Section 2.2.1 of this SE describes additional proposed TS changes. The licensee identified editorial variations where the Sequoyah and Watts Bar TSs use different numbering than the STSs on which TSTF-577 was based. The NRC staff finds that different TS numbering is acceptable because it does not substantively alter TS requirements.
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 NRC staff identified one additional editorial variation. The licensee deleted the word provisions from the Sequoyah, Units 1 and 2, TS 5.5.7 introductory paragraph. The NRC staff has reviewed Sequoyah, Units 1 and 2, TS 5.5.7 and finds that the word, provisions, in the introductory paragraph is duplicative. Therefore, the NRC staff finds the change is acceptable because it does not substantively alter TS requirements.
In addition, the licensee made two additional editorial corrections related to punctuation (adding quotes and a comma) to the Watts Bar, Units 1 and 2, TSs. The NRC staff finds these punctuation changes acceptable because they are consistent with the applicable STS punctuation.
3.2.2 Other Variations Section 2.2.2 of this SE describes additional proposed TS changes. The licensee identified that SQN Unit 1 TSs contains three requirements that differ from the STS on which TSTF-577 was based.
Sequoyah, Unit 1, TS 5.5.7.d.2, and Watts Bar, Unit 1, TS 5.7.2.12d.2, currently permit a permanent SG tube inspection frequency of every 96 EFPM, which is consistent with TSTF-577, and is therefore being retained. The NRC staff finds retaining the 96 EFPM requirement is acceptable because it is consistent with TSTF-577.
Sequoyah, Unit 1, TS 5.5.7.d.2 and Watts Bar, Unit 1, TS 5.7.2.12.d.2, contain a requirement related to inspection probe technology and inspection techniques. This requirement is being deleted for consistency with TSTF-577. The NRC staff finds the change acceptable because the change maintains consistency with TSTF-577.
Sequoyah, Unit 1, TS 5.6.6, and Watts Bar, Unit 1, TS 5.9.9, contain a requirement to discuss trending of tube degradation. This requirement is being deleted for consistency with TSTF-577 and because a similar requirement exists in the revised SQN Unit 1 TS 5.6.6.c. The NRC staff finds the change acceptable because the change maintains consistency with TSTF-577.
Watts Bar, Units 1 and 2, TS 5.7.2.12.b.2 states, Leakage for all degradation mechanisms is not to exceed 150 gpd [gallons per day] for each unfaulted SG. Leakage for all degradation mechanisms is not to exceed 1 gpm in the faulted SG. The Westinghouse STS 5.5.9, Steam Generator (SG) Program, item b.2, contained in TSTF-577, Revision 1, states, Leakage is not to exceed [1 gpm] per SG. The NRC staff notes that the leakage criteria wording in Watts Bar, Units 1 and 2, TS 5.7.2.12.b.2 remains unchanged in this LAR. The NRC staff finds the Watts Bar, Units 1 and 2, TS 5.7.2.12.b.2 wording acceptable because the NRC previously reviewed and approved the leakage criteria for Watts Bar (ML21153A049 (Unit 1) and ML19063B721 and ML21306A287 (Unit 2)) and the criteria remain unchanged in this LAR to adopt TSTF-577.
3.3 TS Change Consistency The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The NRC staff finds that the proposed changes are consistent with Chapter 16.0 of the SRP and are therefore acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendments on September 29, 2022. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change an inspection requirement. The amendments also change reporting requirements. 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 on June 14, 2022, and there has been no public comment on such finding published in the Federal Register (87 FR 36009).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 51.22(c)(10). 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 Contributor: C. Ashley, NRR Date: October 24, 2022
ML22276A161 OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DSS/STSB/BC NAME KGreen RButler VCusumano DATE 09/30/22 10/07/22 09/23/22 OFFICE NRR/DORL/LPLII-2/BC NRR/DORL/LPLII-2/PM NAME DWrona KGreen DATE 10/24/22 10/24/22