ML26023A013
| ML26023A013 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 01/29/2026 |
| From: | Kimberly Green Plant Licensing Branch II |
| To: | Erb D Tennessee Valley Authority |
| Green K | |
| References | |
| EPID L-2024-LLA-0147 | |
| Download: ML26023A013 (0) | |
Text
January 29, 2026 Mr. Delson C. Erb Vice President, OPS Support Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - CORRECTIONS TO AMENDMENT NOS. 177 AND 82 REGARDING USE OF ONLINE MONITORING (EPID L-2024-LLA-0147)
Dear Mr. Erb:
During the issuance of Amendment No. 177 to Facility Operating License (FOL) No. NPF-90 and Amendment No. 82 to FOL No. NPF-96 on December 9, 2025 (ADAMS Accession No. ML25328A120), the U.S. Nuclear Regulatory Commission (NRC) staff inadvertently omitted a hyphen from the word insitu in item a.2 under Watts Bar Nuclear Plant, Unit 1, Technical Specification 5.7.2.24, Online Monitoring Program. The staff also noticed that the word RELAY was misspelled on TS page 1.1-6 of Watts Bar Nuclear Plant, Unit 2, although the definition for SLAVE RELAY TEST was not modified by Amendment No. 82.
The purpose of this letter is to issue corrected TS page 5.0-25c for FOL No. NPF-90 and TS page 1.1-6 for FOL No. NPF-96. The NRC staff has determined that these are typographical errors and do not change the staffs previous conclusion in the safety evaluation for issuance of Amendment Nos. 177 and 82, nor does it affect the NRC staffs no significant hazards consideration determination.
Enclosed, please find corrected TS page 5.0-25c for FOL No. NPF-90 for Amendment No. 177 and corrected TS page 1.1-6 for FOL No. NPF-96 for Amendment No. 82.
If you have any questions regarding this matter, please contact me at (301) 415-1627 or by email at Kimberly.Green@nrc.gov.
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. 50-390 and 50-391
Enclosures:
Corrected TS page 5.0-25c for FOL NPF-90 Corrected TS page 1.1-6 for FOL NPF-96 cc: Listserv
ENCLOSURES WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391 CORRECTED TECHNICAL SPECIFICATION PAGE 5.0-25c FOR FACILITY OPERATING LICENSE NO. NPF-90 CORRECTED TECHNICAL SPECIFICATION PAGE 1.1-6 FOR FACILITY OPERATING LICENSE NO. NPF-96
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 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.
(continued)
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 RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
ML26023A013 NRR-106 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME KGreen ABaxter DWrona KGreen DATE 1/23/2026 01/26/2026 01/29/2026 01/29/2026