ML25328A120
| ML25328A120 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah, Watts Bar  |
| Issue date: | 12/09/2025 |
| From: | Kimberly Green NRC/NRR/DORL/LPL2-2 |
| To: | Erb D Tennessee Valley Authority |
| References | |
| EPID L-2024-LLA-0147 | |
| Download: ML25328A120 (0) | |
Text
December 9, 2025 Mr. Delson C. Erb Vice President, OPS Support Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2; AND WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 373, 368; 177, AND 82 REGARDING USE OF ONLINE MONITORING METHODOLOGY (EPID L-2024-LLA-0147)
Dear Mr. Erb:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 373 to Renewed Facility Operating License No. DPR-77, and Amendment No. 368 to Renewed Facility Operating License No. DPR-79, for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN), respectively; and Amendment No. 177 to Facility Operating License No.
NPF-90, and Amendment No. 82 to Facility Operating License No. NPF-96, for the Watts Bar Nuclear Plant, Units 1 and 2 (WBN), respectively. These amendments are in response to your application dated November 4, 2024.
The amendments revise SQNs and WBNs TS 1.1, Use and Application Definitions, and add new SQN TS 5.5.19 and new WBN TS 5.7.2.24, Online Monitoring Program, to permit the use of an online monitoring methodology (OLM) as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results.
D. Erb A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Kimberly J. Green, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-327, 50-328, 50-390, and 50-391
Enclosures:
- 1. Amendment No. 373 to DPR-77
- 2. Amendment No. 368 to DPR-79
- 3. Amendment No. 177 to NPF-90
- 4. Amendment No. 82 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. 373 Renewed License No. DPR-77 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated November 4, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR)
Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-77 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 373 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 6 months of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: December 9, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.12.09 09:13:25 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 373 SEQUOYAH NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace page 3 of the Renewed Facility Operating License with the attached page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following pages of Appendix A, Technical Specifications, with the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 1.1-1 1.1-1 1.1-3 1.1-3 1.1-5 1.1-5 5.5-18 5.5-18 5.5-19 (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. 373 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. 373 Renewed License No. DPR-77
Definitions 1.1 SEQUOYAH - UNIT 1 1.1-1 Amendment 334, 347, 373 1.0 USE AND APPLICATION 1.1 Definitions
NOTE-----------------------------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.
AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux signals (AFD) between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY (excluding transmitters in the Online Monitoring Program). Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
Definitions 1.1 SEQUOYAH - UNIT 1 1.1-3 Amendment 334, 352, 362, 373 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 (including transmitters in the Online Monitoring Program), 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.
Definitions 1.1 SEQUOYAH - UNIT 1 1.1-5 Amendment 334, 352, 373 1.1 Definitions QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3455 MWt.
REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay.
The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
Programs and Manuals 5.5 SEQUOYAH - UNIT 1 5.5-18 Amendment 358, 373 5.5 Programs and Manuals 5.5.18 Risk Informed Completion Time Program (continued) 2.
For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.
3.
Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.
d.
For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:
1.
Numerically accounting for the increased possibility of CCF in the RICT calculation; or 2.
Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.
e.
The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.
5.5.19 Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.
The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version). The program shall include the following elements:
Programs and Manuals 5.5 SEQUOYAH - UNIT 1 5.5-19 Amendment 373 5.5 Programs and Manuals 5.5.19 Online Monitoring Program (continued) a.
Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.
1.
Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration
- check, 2.
Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance, 3.
Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and 4.
Documentation of the results of the online monitoring data analysis.
b.
Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.
c.
Performance of calibration checks for transmitters at the specified backstop frequencies.
d.
The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 368 Renewed License No. DPR-79 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated November 4, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-79 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 368 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 6 months of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: December 9, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.12.09 09:13:56 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 368 SEQUOYAH NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace page 3 of the Renewed Facility Operating License with the attached page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following pages of Appendix A, Technical Specifications, with the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 1.1-1 1.1-1 1.1-3 1.1-3 1.1-5 1.1-5 5.5-18 5.5-18 5.5-19 (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. 368 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. 368 Renewed License No. DPR-79
Definitions 1.1 SEQUOYAH - UNIT 2 1.1-1 Amendment 327, 341, 368 1.0 USE AND APPLICATION 1.1 Definitions
NOTE-----------------------------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.
AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux signals (AFD) between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY (excluding transmitters in the Online Monitoring Program). Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
Definitions 1.1 SEQUOYAH - UNIT 2 1.1-3 Amendment 327, 346, 356, 368 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 (including transmitters in the Online Monitoring Program), 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.
Definitions 1.1 SEQUOYAH - UNIT 2 1.1-5 Amendment 327, 346, 368 1.1 Definitions QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3455 MWt.
REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay.
The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
Programs and Manuals 5.5 SEQUOYAH - UNIT 2 5.5-18 Amendment 352, 368 5.5 Programs and Manuals 5.5.18 Risk Informed Completion Time Program (continued) 2.
For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.
3.
Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.
d.
For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:
1.
Numerically accounting for the increased possibility of CCF in the RICT calculation; or 2.
Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.
e.
The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.
5.5.19 Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.
The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version). The program shall include the following elements:
Programs and Manuals 5.5 SEQUOYAH - UNIT 2 5.5-19 Amendment 368 5.5 Programs and Manuals 5.5.19 Online Monitoring Program (continued) a.
Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.
1.
Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration
- check, 2.
Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance, 3.
Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and 4.
Documentation of the results of the online monitoring data analysis.
b.
Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.
c.
Performance of calibration checks for transmitters at the specified backstop frequencies.
d.
The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 177 License No. NPF-90 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated November 4, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-90 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 177 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 6 months of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: December 9, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.12.09 09:14:32 -05'00'
ATTACHMENT TO AMENDMENT NO. 177 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 Appendix A, Technical Specifications, with the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Pages Insert Pages 1.1-1 1.1-1 1.1-3 1.1-3 1.1-5 1.1-5 5.0-25c 5.0-25c 5.0-25d
Facility License No. NPF-90 Amendment No. 177 (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. 177 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 Watts Bar-Unit 1 1.1-1 (continued)
Amendment 156, 177 1.0 USE AND APPLICATION 1.1 Definitions
NOTE------------------------------------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term Definition ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)
CHANNEL CALIBRATION ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.
AFD shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. THE CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY (excluding transmitters in the Online Monitoring Program). Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
Definitions 1.1 1.1 Definitions (continued)
(continued)
Watts Bar-Unit 1 1.1-3 Amendment 24, 141, 146, 159, 177 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME La LEAKAGE The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
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 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)
(continued)
Watts Bar - Unit 1 1.1-5 Amendment 24, 31, 141, 177 PHYSICS TESTS a.
Described in Chapter 14, Initial Test Program of the (continued)
FSAR; b.
Authorized under the provisions of 10 CFR 50.59; or c.
Otherwise approved by the Nuclear Regulatory Commission.
PRESSURE AND TEMPERATURE LIMITS REPORT QUADRANT POWER TILT RATIO (QPTR)
RATED THERMAL POWER (RTP)
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME SHUTDOWN MARGIN (SDM)
The PTLR is the unit specific document that provides the RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.9.6. Plant operation within these operating limits is addressed in LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Cold Overpressure Mitigation System (COMS)."
QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage.
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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
Procedures, Programs, and Manuals 5.7 Watts Bar-Unit 1 5.0-25c Amendment 132, 177 5.7 Procedures, Programs, and Manuals 5.7.2.23 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
a.
The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
b.
Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
c.
The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
5.7.2.24 Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.
The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version). The program shall include the following elements:
a.
Implementation of online monitoring for transmitters that have been evaluated in accordance with a NRC approved methodology during the plant operating cycle.
1.
Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check, 2.
Performance of online monitoring using noise analysis to assess insitu dynamic response of transmitters that can affect response time performance, 3.
Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and 4.
Documentation of the results of the online monitoring data analysis.
b.
Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.
(continued)
Procedures, Programs, and Manuals 5.7 Watts Bar-Unit 1 5.0-25d 5.7 Procedures, Programs, and Manuals 5.7.2.24 Online Monitoring Program (continued) c.
Performance of calibration checks for transmitters at the specified backstop frequencies.
d.
The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.
Amendment 177
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 82 License No. NPF-96
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated November 4, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-96 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 82 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 6 months of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: December 9, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.12.09 09:15:02 -05'00'
ATTACHMENT TO AMENDMENT NO. 82 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 Appendix A, Technical Specifications, with the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Pages Insert Pages 1.1-1 1.1-1 1.1-3 1.1-3 1.1-6 1.1-6 5.0-27c Unit 2 Facility Operating License No. NPF-96 Amendment No. 82 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. 82 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 Watts Bar - Unit 2 1.1-1 (continued)
Amendment 64, 82 1.0 USE AND APPLICATION 1.1 Definitions
NOTE----------------------------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term Definition ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)
CHANNEL CALIBRATION ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.
AFD shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY (excluding transmitters in the Online Monitoring Program). Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
Definitions 1.1 1.1 Definitions (continued)
Watts Bar - Unit 2 1.1-3 (continued)
Amendment 47, 52, 82 DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME La DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.
The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
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).
Definitions 1.1 1.1 Definitions (continued)
Watts Bar - Unit 2 1.1-6 (continued)
Amendment 42, 47, 64, 82 QUADRANT POWER TILT RATIO (QPTR)
RATED THERMAL POWER (RTP)
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE REALY TEST may be performed by means of any series of sequential, overlapping, or total steps.
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals Watts Bar - Unit 2 5.0-27c Amendment 82 5.7.2.24 Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.
The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version). The program shall include the following elements:
a.
Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.
1.
Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check, 2.
Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance, 3.
Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and 4.
Documentation of the results of the online monitoring data analysis.
b.
Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.
c.
Performance of calibration checks for transmitters at the specified backstop frequencies.
d.
The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 373 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 AMENDMENT NO. 368 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 AMENDMENT NO. 177 TO FACILITY OPERATING LICENSE NO. NPF-90 AMENDMENT NO. 82 TO FACILITY OPERATING LICENSE NO. NPF-96 SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391 TENNESSEE VALLEY AUTHORITY
1.0 INTRODUCTION
By application dated November 4, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML24309A061), the Tennessee Valley Authority (TVA or the licensee) requested changes to the technical specifications (TSs) for Sequoyah Nuclear Plant, Units 1 and 2 (SQN), and Watts Bar Nuclear Plant, Units 1 and 2 (WBN).
The proposed amendments would revise some definitions in SQNs and WBNs TS 1.1, Definitions, and add new TS 5.5.19, Online Monitoring Program, (SQN) and add new TS 5.7.2.24, Online Monitoring Program (WBN), to permit the use of a U.S. Nuclear Regulatory Commission (NRC or Commission)-approved online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The proposed amendments are based on the NRC approved topical report AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (ML21235A493).
The NRC staff issued a safety evaluation (SE) approving the -A version of the AMS-TR-0720R2-A on August 11, 2021, Final Safety Evaluation for AMS [Analysis and Measurement Services] Online Monitoring Topical Report (package ML21179A060). The SE states, in part, that the NRC staff finds that implementation of an OLM program in accordance with the approved AMS OLM TR provides an acceptable alternative to periodic manual calibration surveillance requirements upon implementation of the application-specific action items. TVA has not proposed any deviations from the approved AMS-TR-0720R2-A.
The NRC staff conducted a virtual regulatory audit between June 17 and August 7, 2025, to examine the licensees non-docketed information on the proposed SQN and WBN OLM methodologies. The NRC staff did not identify any need for additional information and issued an audit summary dated November 20, 2025 (ML25272A158).
2.0 REGULATORY EVALUATION
2.1 Background
The transmitters that the licensee plans to include in the OLM program provide input to the reactor trip system (RTS) and engineered safety feature actuation systems (ESFAS) and are used for post accident monitoring (PAM), the remote shutdown system (RSS), and low temperature overpressure protection (LTOP) or cold overpressure mitigation system (COMS).
The RTS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and reactor coolant system (RCS) pressure boundary during anticipated operational occurrences and to assist the engineered safety feature (ESF) systems in mitigating accidents. The RTS and related instrumentation are identified in the SQN and WBN TS Table 3.3.1-1.
The ESFAS and related systems initiate necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the RCS pressure boundary, and to mitigate accidents. The ESFAS and related instrumentation are identified in the SQN and WBN TS Table 3.3.2-1.
The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design-basis accidents. The PAM instrumentation is identified in the SQN and WBN TS Table 3.3.3-1.
The RSS provides the operator with sufficient instrumentation and controls to place and maintain the unit in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility that the control room becomes inaccessible. The RSS instrumentation is addressed in the SQN and WBN TS 3.3.4 and identified in SQN and WBN Bases Table B 3.3.4-1.
The LTOP (SQN) or COMS (WBN) controls RCS overpressure at low temperatures, so the integrity of the reactor coolant pressure boundary is not compromised by violating the pressure and temperature limits of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix G, Fracture Toughness Requirements. The LTOP or COMS instrumentation is identified in SQN and WBN TS 3.4.12, respectively.
Switching from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency will not create any physical changes to the plants, and the changes will not impact how the plants operate. The licensee stated that it will use condition-based frequency to determine when transmitter calibrations are needed instead of performing calibrations based on a calendar frequency. The licensee plans to use existing calibration methods when it is determined that transmitter calibration is needed.
2.2 Requested Changes As described in the license amendment request (LAR) dated November 4, 2024, the licensee proposed changes to the following definitions in SQN TS 1.1, Use and Application Definitions.
Additions of text are shown in bolded, underlined text.
CHANNEL CALIBRATION would be revised as follows:
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY (excluding transmitters in the Online Monitoring Program).
Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME would be revised as follows:
The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME would be revised as follows:
The RTS RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
As described in the LAR, the licensee proposed changes to the following definitions in WBN TS 1.1, Use and Application Definitions.
CHANNEL CALIBRATION would be revised as follows:
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY (excluding transmitters in the Online Monitoring Program).
Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME would be revised as follows:
The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME would be revised as follows:
The RTS RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. 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 (including transmitters in the online monitoring program), or the components have been evaluated in accordance with an NRC approved methodology.
The licensee proposed to add new TS 5.5.19, Online Monitoring Program, (SQN) and new TS 5.7.2.24, Online Monitoring Program (WBN) as follows:
Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.
The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version). The program shall include the following elements:
- a. Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.
- 1. Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration
- check,
- 2. Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,
- 3. Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and
- 4. Documentation of the results of the online monitoring data analysis.
- b. Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.
- c. Performance of calibration checks for transmitters at the specified backstop frequencies.
- d. The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.
2.3 Applicable Regulatory Requirements and Guidance The NRC staff considered the following regulatory requirements and guidance in reviewing the licensees request:
The regulations at 10 CFR 50.36(c)(1)(ii)(A) requires, in part, that limiting safety system settings (LSSS) are settings for automatic protective devices related to those variables having significant safety functions. Where an LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded. It also requires that the licensee take appropriate action and notify the NRC if the licensee determines that an automatic safety system does not function as required. The licensee is then required to notify the Commission, review the matter and record the results of the review.
The regulation at 10 CFR 50.36(c)(3) states, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
The regulation at 10 CFR 50.36(c)(5) states, in part, 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.
Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, establishes the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission.
As noted in the LAR, both SQN and WBN were designed to meet the intent of the Proposed General Design Criteria for Nuclear Power Plant Construction Permits, published in July 1967.
The SQN construction permit was issued in May 1970, and the WBN construction permit was issued in January 1973. Each plants UFSAR addresses the NRC GDC published as Appendix A to 10 CFR Part 50 in July 1971.
The NRC determined that the following GDC are applicable to this review:
GDC 13, Instrumentation and control, states that [i]nstrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
GDC 20, Protection system functions, states that [t]he protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
The following are the specific NRC guidance documents applicable to the NRC staffs evaluation of the SQN and WBN OLM programs:
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Branch Technical Position (BTP) 7-12, Guidance on Establishing and Maintaining Instrument Setpoints, Revision 6, August 2016 (ML16019A200)
Regulatory Guide (RG) 1.105, Revision 4, Setpoints for Safety-Related Instrumentation,,
February 2021 (ML20330A329). This RG describes an approach that is acceptable to the NRC staff to meet regulatory requirements to ensure that: (a) setpoints for safety-related instrumentation are established to protect nuclear power plant safety and analytical limits, and (b) the maintenance of instrument channels implementing these setpoints ensures they are functioning as required, consistent with the plant TSs.
This RG endorses American National Standards Institute (ANSI)/International Society of Automation (ISA) Standard 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation. Among other things, the ANSI/ISA 67.04.01 standard provides criteria for assessing the performance of safety related instrument channels to ensure they remain capable of achieving their required safety functions in a reliable manner. This performance monitoring process requires the establishment of acceptable As-Found tolerance limits used to check whether an instrument channel is functioning as required, and the establishment of acceptable As-Left tolerance limits used to establish the maximum allowed deviation from the desired setpoint of the instrument channel and still be considered as within calibration.
3.0 TECHNICAL EVALUATION
3.1 Description of the OLM Program The proposed SQN and WBN OLM programs are based on the AMS OLM TR, AMS-TR-0720R2-A, which provides a methodology for performing OLM of the output signals of pressure and differential pressure transmitters. This methodology was developed by AMS to be used in nuclear power plants as an analytical tool to monitor sensor deviations from proper calibration conditions by monitoring performance of the sensor during plant operation between scheduled refueling outages. The purpose of this monitoring is to flag to plant personnel any sensor performance that has deviated sufficiently from ideal calibration conditions to warrant a required calibration to be performed for that channel at the next refueling outage.
3.1.1 OLM Program Implementation The licensee stated in section 3.2 of its LAR, that the AMS Bridge and the AMS Calibration Reduction System software programs were developed under AMSs 10 CFR Part 50, Appendix B, compliant Quality Assurance (QA) program. The NRC staff conducted an inspection of AMS to review AMSs implementation of its QA program with respect to the design, testing, and error controls for the AMS Bridge and the AMS Calibration Reduction System software programs. The NRC staff documented its inspection findings in an inspection report dated March 14, 2025, Nuclear Regulatory Commission Inspection Report of Analysis and Measurement Services No. 99902075/2025-201 (ML25071A181). As stated in this inspection report, the NRC staff determined that AMS is implementing its design control and test control program in accordance with the regulatory requirements of Criterion III, Design Control, and Criterion XI, Test Control, of Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50, for the AMS Bridge and the AMS Calibration Reduction System software programs. In addition, the NRC staff reviewed AMSs processes for controlling software errors and with exception of one minor procedural issue, the NRC staff determined that AMS is implementing its non-conforming materials, parts, or components program in accordance with the regulatory requirements of Criterion XV, Nonconforming Materials, Parts, or Components, of Appendix B to 10 CFR Part 50 for the AMS Bridge and the AMS Calibration Reduction System software programs.
3.2 Description and Evaluation of TS Changes The licensees submittal requested approval to implement its OLM program by revising appropriate TS 1.1, Definitions, and adding new sections titled, Online Monitoring Program, for SQN, TS 5.5.19, and for WBN, TS 5.7.2.24. The licensee proposes to use the OLM methodology presented in AMS-TR-0720R2-A as the technical basis to change from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM analysis results. Markup of the TS pages were provided in attachments 1 (SQN) and 2 (WBN) of the enclosure to the LAR.
The regulation at 10 CFR 50.36(a)(1) states, in part, that: [a] summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application but shall not become part of the technical specifications.
Accordingly, the licensee also submitted, for NRC staff information, the proposed TS Bases changes that correspond to the proposed TS changes. The NRC staff did not review the TS Bases.
3.3 OLM Noise Analysis Implementation The licensee provided as part of its LAR, the steps to implement the noise analysis technique to assess dynamic failure modes of pressure transmitters. The information provided in sections 3.3.1 through 3.3.6 of the LAR is mapped to the steps found in section 11.3.3, Steps for Implementation of Noise Analysis Technique, of AMS-TR-0720R2-A. The NRC staff finds that the provided mapping and information related to the noise analysis implementation is consistent with the AMS OLM TR, AMS-TR-0720R2-A.
3.4 TS 1.1, Use and Application Definitions For both SQN, Units 1 and 2, and WBN, Units 1 and 2, the TS definition for the term CHANNEL CALIBRATION is being revised to account for the approved OLM methodologies. The specific change allows transmitters that are included in the licensees OLM program to be excluded from the scope of instrumentation to be periodically calibrated within the frequency in the Surveillance Frequency Control Program (SFCP).
The NRC staff reviewed this proposed change considering the context of the OLM program.
This change is acceptable because the OLM processes would include an acceptable method for identifying performance issues as they occur and initiating corrective actions when preestablished OLM limits are exceeded. The corrective actions would also include performing instrument calibrations as necessary to restore instrument performance to within acceptable parameters. Data collected during OLM activities would be used to adjust OLM limits such that poorly performing instruments would be calibrated at greater frequencies to address any potential impact on long term plant performance.
For both SQN and WBN, the TS definition for the terms ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME and REACTOR TRIP SYSTEM (RTS) RESPONSE TIME are being revised to extend the current exclusion from periodic response time testing for instruments that are entered into the OLM program. The previous exclusion from response time testing had been based on the periodic channel calibration program, which will be replaced with the OLM program for those instruments that are included in the OLM scope.
The NRC staff finds these revised definitions to be acceptable because the OLM program will continue to monitor instrument performance and will be capable of detecting instrument degradation or failures that could affect response time performance. The previous definitions for these terms allowed exclusion from response time testing because instrument failures that affect time response would be detectable during the periodic calibration tests and channel check activities. Since the OLM program will retain the capability of detecting and correcting instrument degraded performance or fault conditions, the NRC staff considers this method to be an acceptable and approved methodology to support continued exception of these instruments from time response testing.
3.5 New Online Monitoring Program for SQN, TS 5.5.19 and WBN, TS 5.7.2.24 New SQN, TS 5.5.19 and WBN, TS 5.7.2.24 provide a description of the AMS-based OLM program, including key elements. The new TS stipulates that the OLM program must be implemented in accordance with the NRC-approved AMS OLM TR, AMS-TR-0720R2-A. The NRC staff reviewed the TS description of the OLM program in the LAR and found that it is consistent with the program description provided in the approved AMS OLM TR, AMS-TR-0720R2-A. To verify that the SQN and WBN programs will be implemented in accordance with the NRC-approved TR, the NRC staff conducted an audit per an audit plan dated February 27, 2025 (ML25052A088), and examined several SQN and WBN specific reports that documented program implementation activities. These reports are described in the NRC staffs audit summary report dated November 20, 2025 (ML25272A158). The NRC staff audit confirmed that key elements including calculations of OLM limits, amenable transmitters to be included in the OLM program, backstop calculations, noise analysis implementation, maximum sampling rate calculations, OLM coverage of transmitter setpoints and range, drift monitoring, plant procedures for data retrieval, and analysis and capture of data for the OLM program would be implemented as described in AMS-TR-0720R2-A.
The NRC staff also reviewed TVAs responses to each of the Application Specific Action Items (ASAIs) that are contained in section 4.0 of the NRC SE for AMS OLM TR, AMS-TR-0720R2-A.
These licensee responses are provided in section 3.4 of the LAR dated November 4, 2024. The NRC staff evaluated these ASAIs in section 3.6 of this SE. The NRC determined that all plant-specific actions and the SQN and WBN OLM programs would be implemented in conformance with the approved AMS OLM TR, AMS-TR-0720R2-A.
3.6 AMS TR-0720R2-A - ASAIs 3.6.1 ASAI 1 - Evaluation and Proposed Mark-up of Existing Plant Technical Specifications ASAI 1:
When preparing a license amendment request to adopt OLM methods for establishing calibration frequency, licensees should consider markups that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance. Such TS changes would need to include appropriate markups of the TS tables describing limiting conditions for operation and surveillance requirements, the technical basis for the changes, and the administrative programs section.
The licensee provided markups of the applicable TSs that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance for transmitters that are included in the OLM program. Markups of the TS Bases were also provided for information only, which describe the technical basis for the OLM program. Therefore, the NRC staff finds that ASAI 1 is met.
3.6.2 ASAI 2 - Identification of Calibration Error Source ASAI 2:
When determining whether an instrument can be included in the plant OLM program, the licensee shall evaluate calibration error source in order to account for the uncertainty due to multiple instruments used to support the transfer of transmitter signal data to the data collection system. Calibration errors identified through OLM should be attributed to the transmitter until testing can be performed on other support devices to correctly determine the source of calibration error and reallocate errors to these other loop components.
The NRC staff performed an audit of the SQN and WBN OLM program reports to verify that calibration error sources were being factored into account for the uncertainty due to multiple instruments used to support the transfer of transmitter signal data to the data collection system.
The NRC staff confirmed that the OLM program attributes calibration errors to the transmitter unless testing is subsequently performed to determine and reallocate calibration error to other instrument loop components. Therefore, the NRC staff finds that ASAI 2 is met.
3.6.3 ASAI 3 - Response Time Test Elimination Basis ASAI 3:
If the plant has eliminated requirements for performing periodic RT [response time] testing of transmitters to be included in the OLM program, then the licensee shall perform an assessment of the basis for RT test elimination to determine if this basis will remain valid upon implementation of the OLM program and to determine if the RT test elimination will need to be changed to credit the OLM program rather than the periodic calibration test program.
The transmitters that are being incorporated into the OLM program were excluded from RT testing. The licensee, therefore, performed an assessment of the basis for RT testing exclusions and determined that the OLM program will continue to support exclusion from RT testing because the OLM methods will detect transmitter failures that would affect RT performance. The basis for this exclusion in TS 1.1 is evaluated in section 3.4 of this SE. Therefore, the NRC staff finds that ASAI 3 is met.
3.6.4 ASAI 4 - Use of Calibration Surveillance Interval Backstop ASAI 4:
In its application for a license or license amendment to incorporate OLM methods for establishing calibration surveillance intervals, applicants or licensees should describe how they intend to apply backstop intervals as a means for mitigating the potential that a process group could be experiencing undetected common mode drift characteristics.
The NRC staff performed an audit which included the backstop interval calculations performed for both SQN and WBN transmitters being incorporated into the proposed OLM program and confirmed that these calculations were performed in a manner consistent with the processes outlined in the approved AMS OLM TR for determining maximum calibration intervals.
Therefore, the NRC staff finds that ASAI 4 is met.
3.6.5 ASAI 5 - Use of Criteria other than in AMS OLM TR for Establishing Transmitter Drift Flagging Limit ASAI 5:
In its application for a license or license amendment to incorporate OLM methods for establishing calibration surveillance intervals, applicants or licensees should describe whether they intend to adopt the criteria within the AMS OLM TR for flagging transmitter drift or whether they plan to use a different methodology for determining this limit.
The NRC staff determined that the SQN and WBN proposed OLM programs are consistent with the AMS OLM TR, AMS-TR-0720R2-A, and therefore, a different methodology is not being employed. Therefore, the NRC staff finds that ASAI 5 is met.
3.7 Technical Summary The NRC staff finds that the licensees proposed implementation of the SQN and WBN OLM programs are consistent with the approved AMS OLM TR, AMS-TR -720R2-A. The NRC staff also finds the proposed revision to each units TS 1.1 and proposed addition of SQN TS 5.5.19, and WBN TS 5.7.2.24 acceptable.
The NRC staff determined that implementation of the proposed OLM program for SQN and WBN will continue to support establishment and maintenance of LSSS associated with the transmitters that are included in the program. These settings will continue to ensure that associated automatic protective actions will correct abnormal situations before safety limits are exceeded. Implementation of the OLM programs at SQN and WBN would identify those protection system instrument channels that require recalibration, which helps to ensure that the licensee would take appropriate actions if the licensee determines that an automatic safety system does not function as required. The surveillance requirements relating to test, calibration, and inspection of these transmitters will also continue to ensure that the adequate quality of systems and components is maintained. Therefore, the NRC staff finds that the requirements of 10 CFR 50.36(c)(1)(ii)(A), and 10 CFR 50.36(c)(3) will continue to be met.
Further, 10 CFR 50.36(c)(5) is met by the addition of the new program to the licensees TSs.
Additionally, the NRC staff finds that the licensees implementation of the OLM Program in accordance with approved TR AMS-TR-720R2-A will continue to meet the requirements of principal design criteria GDCs 13 and 20 for both SQN, Units 1 and 2, and WBN, Units 1 and 2, as documented in their respective UFSARs. Per 10 CFR 50.36(c)(1)(ii)(A), the licensee is required to notify the NRC if an associated automatic safety system does not function as required.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendments on November 21, 2025. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change surveillance 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 published in the Federal Register on January 21, 2025 (90 FR 7190), and there has been no public comment on such finding.
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: G. Blas, NRR F. OBrien, NRR D. Rahn, NRR T. Sweat, NRR A. Armstrong, NRR Date: December 9, 2025
ML25328A120 NRR-058 OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DEX/EICB/BC(A)
NRR/DRO/IQVB/BC NAME KGreen ABaxter (PBlechman for)
SDarbali KKavanagh DATE 11/24/2025 12/5/2025 9/29/2025 12/4/2025 OFFICE NRR/DSS/STSB/BC NRR/DORL/LPLII-2/BC NRR/DORL/LPLII-2/PM NAME SMehta DWrona KGreen DATE 12/5/2025 12/9/2025 12/9/2025