IR 05000390/2025003
| ML25343A278 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 12/17/2025 |
| From: | Renee Taylor NRC/RGN-II/DORS/PB5 |
| To: | Erb D Tennessee Valley Authority |
| References | |
| EA-NMSS-2023-0002, EAF-NMSS-2025-0215 IR 2025003, IR 2024002 | |
| Download: ML25343A278 (0) | |
Text
SUBJECT:
WATTS BAR NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000390/2025003, 05000391/2025003, AND 07201048/2024002 AND EXERCISE OF ENFORCEMENT DISCRETION
Dear Delson Erb:
On September 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Watts Bar Nuclear Plant. On December 8, 2025, the NRC inspectors discussed the results of this inspection with Chris Reneau, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
Due to the temporary cessation of government operations, which commenced on October 1, 2025, the NRC began operating under its Office of Management and Budget-approved plan for operations during a lapse in appropriations. Consistent with that plan, the NRC operated at reduced staffing levels throughout the duration of the shutdown. However, the NRC continued to perform critical health and safety functions and make progress on other high-priority activities associated with the ADVANCE Act and Executive Order 14300. On November 13, 2025, following the passage of a continuing resolution, the NRC resumed normal operations. However, due to the 43-day lapse in normal operations, the Office of Nuclear Reactor Regulation granted the Regional Offices an extension on the issuance of the calendar year 2025 inspection reports that should have been issued by November 13, 2025, to December 31, 2025. The NRC resumed the routine cycle of issuing inspection reports on November 13, 2025.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
In addition, the NRC identified a violation of Title 10 of the Code of Federal Regulations (CFR)
72.48, paragraphs (c)(1), (c)(2), and (d)(1), and provisions of 10 CFR 72.212 that resulted from a certificate of compliance (CoC) holders failure to comply with 10 CFR 72.48 for a CoC holder-generated design change to its multi-purpose canister fuel basket, known as the continuous basket shim variant, which altered the structural configuration from welded to bolted shims.
However, an Interim Enforcement Policy (IEP) issued in August 2025 is applicable to this violation. Specifically, Enforcement Policy Section 9.4, Enforcement Discretion for General December 17, 2025 Licensee Adoption of Certificate of Compliance Holder-Generated Modifications under 10 CFR Part 72.48, provides enforcement discretion to not issue an enforcement action for this violation. The licensee will be expected to comply with 10 CFR 72.212 provisions after the NRC dispositions the noncompliance for a CoC holder-generated change that affects the General Licensee.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Watts Bar Nuclear Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Watts Bar Nuclear Plant.
This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Ryan C. Taylor, Chief Projects Branch 5 Division of Operating Reactor Safety Docket Nos. 05000390, 05000391, and 07201048 License Nos. NPF-90 and NPF-96
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000390, 05000391, and 07201048
License Numbers:
Report Numbers:
05000390/2025003, 05000391/2025003, and 07201048/2024002
Enterprise Identifier:
I-2025-003-0027 and I-2024-002-0078
Licensee:
Tennessee Valley Authority
Facility:
Watts Bar Nuclear Plant
Location:
Spring City, Tennessee
Inspection Dates:
July 1, 2025 to September 30, 2025
Inspectors:
P. Cooper, Senior Reactor Inspector
S. Hamilton, Resident Inspector
D. Hardage, Senior Resident Inspector
C. Kline, Senior Resident Inspector
D. Neal, Health Physicist
A. Price, Resident Inspector
R. Wehrmann, Senior Resident Inspector
Approved By:
Ryan C. Taylor, Chief
Projects Branch 5
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Watts Bar Nuclear Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Demonstrate Effective Control of Performance of a Maintenance Rule System Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000390,05000391/2025003-01 Open/Closed
[P.1] -
Identification 71111.12 NRC inspectors identified a Green finding and associated non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants," paragraph (a)(2), for the licensee's failure to demonstrate effective control of performance of the Watts Bar unit 2 main steam valve vault room ventilation system through performance of appropriate preventive maintenance.
Incorrect Rod Control Setup Resulted in Unanticipated Control Rod Withdrawal Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000391/2025003-02 Open/Closed
[H.11] -
Challenge the Unknown 71111.12 NRC inspectors identified a Green finding and associated non-cited violation of Watts Bar Unit 2 Technical Specification 5.7.1.1a, "Procedures," for the licensee's failure to ensure that appropriate instructions were established in a plant startup procedure associated with the control rod drive system. This directly resulted in improper setup of the control rod bank overlap circuitry for the unit 2 control rod banks and resulted in unanticipated and unplanned withdrawal of control rod bank B.
Additional Tracking Items
Type Issue Number Title Report Section Status EDG EAF-NMSS-2025-0215 Interim Enforcement Policy (IEP) Associated with the Continuous Basket Shim 60855 Closed
PLANT STATUS
Unit 1 was at or near rated thermal power for the entire inspection period.
Unit 2 began the inspection period at rated thermal power. On July 13, 2025, the unit tripped due to a loss of both turbine-driven main feedwater pumps. The reactor was restarted on July 16, 2025, and returned to rated thermal power on July 18, 2025, where it remained at or near for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed onsite portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal hot temperatures for the following systems:
- auxiliary building ventilation
- turbine-driven auxiliary feedwater pumps
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)unit 1 A-A emergency diesel generator on July 14, 2025 (2)unit 2 A-A safety injection pump on July 23, 2025
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a complete walkdown of the unit 2 A-A containment spray system on August 27, 2025.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
(1)unit 1 pipe chases elevation 676', 692', and 713' on July 7, 2025 (2)unit 2 safety injection pump rooms on July 22, 2025 (3)unit 1 residual heat removal pump rooms on July 22, 2025 (4)unit 2 pressurizer heater transformer room and control rod drive motor generator set room on July 23, 2025
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on July 8, 2025.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during unit 2 startup on July 16, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator requalification as-found simulator sets for cycle 25-02 (crew 1) on July 1, 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
(1)control banks A and B withdrew at the same time during unit 2 reactor startup, under condition report (CR) 2026418 and work order (WO) 125479975 on July 16, 2025 (2)south valve vault room HVAC, under CRs 2028251, 2025041, 2025581, and 2020218
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
(1)unit 1 and unit 2, week of July 20-26, including protected equipment reviews for scheduled maintenance on 2B emergency diesel generator (2)unit 1 and unit 2, week of August 11-14, due to emergent repairs to the fire water header in the auxiliary building (3)unit 1 and unit 2, week of August 24-28, due to emergent repairs to the 1B traveling water screen
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1)emergency diesel generator battery not meeting surveillance acceptance criteria, as documented in CR 2025000 on July 9, 2025
- (2) VLF test results on 2B emergency diesel generator B phase cable displays a degrading trend, as documented in CR 2027558 on July 21, 2025 (3)2-FCV-043-0202-A, loss of coolant accident hydrogen containment monitor isolation valve, failed to meet acceptance criteria per 2-SI-43-901-A, as documented in CR 2033758 on August 20, 2025 (4)1-FCV-67-22, emergency raw cooling water strainer 1A-A inlet isolation valve, could not be stroked from the main control room, as documented in CR 2034073 on August 22, 2025 (5)amber light for steam generator #4 main steam isolation valve test switch was lit with the hand switch in the normal position, as documented in CR 2033216 on August 26, 2025 (6)engineering review for installed item DAN140R, MCC cubicle, with applicable part 21 event # 55223, as documented in CR 2033552
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (5 Samples)
(1)1B-B emergency diesel generator post maintenance testing on July 10, 2025, under WO 124971241 (2)2B-B emergency diesel generator post maintenance testing on July 24, 2025, under WO 124424699 (3)1A-A auxiliary building gas treatment fan on July 28, 2025, under WO 124664943
- (4) A-A emergency raw cooling water pump on July 30, 2025, under WO 124923301
- (5) A-A emergency raw cooling water pump after motor replacement on August 22, 2025, under WO 123085887
Surveillance Testing (IP Section 03.01) (1 Sample)
(1)1A-A auxiliary building gas treatment system filter testing per 0-SI-30-9-A on July 30, 2025, under WO 123572953
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)unit 2 containment spray pump 2A-A comprehensive test per 0-SI-72-908-A on August 1, 2025, under WO 124969931
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
On September 3, 2025, the inspectors evaluated an emergency preparedness drill which included the following:
(1)an anticipated transient without a scram, alert declaration, loss of coolant accident, containment spray pump trip, loss of reactor vessel level, and general emergency declaration
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS06: Emergency AC Power Systems (IP Section 02.05) (2 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
- (2) unit 2 (July 1, 2024, through June 30, 2025)
MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
- (2) unit 2 (July 1, 2024, through June 30, 2025)
MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
- (2) unit 2 (July 1, 2024, through June 30, 2025)
MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
- (2) unit 2 (July 1, 2024, through June 30, 2025)
MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
- (2) unit 2 (July 1, 2024, through June 30, 2025)
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (2 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
- (2) unit 2 (July 1, 2024, through June 30, 2025)
BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
(2)unit 2 (July 1, 2024, through June 30, 2025) 71152A - Annual Follow-up Problem Identification and Resolution
Annual Follow-up of Selected Issues (Section 03.03) (1 Sample)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
(1)complete review of all past performances of preventive maintenance activity 600124697, "Periodic Testing of Plant Public Address System," documented in CRs 2023435 and 2021575 71153 - Follow-Up of Events and Notices of Enforcement Discretion
Personnel Performance (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated an uncomplicated reactor trip on unit 2 due to a turbine trip, and the licensees performance on July 13, 2025.
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL 60855 - Operation of an Independent Spent Fuel Storage Installation Inspections were conducted using the appropriate portions of the IPs in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with IMC 2690, Inspection Program for Storage of Spent Reactor Fuel and Reactor-Related Greater-than-Class C Waste at Independent Spent Fuel Storage Installations (ISFSI) and for 10 CFR Part 71 Transportation Packagings.
Operation of an Independent Spent Fuel Storage Installation (1 Sample)
- (1) The inspector conducted a periodic in-office follow-up that focused on the review of the licensees implementation of the 10 CFR 72.48 process and associated corrective actions related to ISFSI activities. The review included:
- 72.48 evaluations and screenings: reviewed the licensees 72.48 process and associated evaluation associated with the adoption of the continuous basket shim (CBS) basket variant
- corrective action program: reviewed condition reports related to the design change of the CBS basket variant
INSPECTION RESULTS
Enforcement Discretion Enforcement Action EAF-NMSS-2025-0215: Interim Enforcement Policy (IEP) associated with the Continuous Basket Shim 60855
Description:
Holtec International (also referred to as the certificate of compliance (CoC)holder) implemented a design change to its multi-purpose canister fuel basket, known as the continuous basket shim variant, which altered the structural configuration from welded to bolted shims. This change resulted in a departure from the method of evaluation (MOE)described in the final safety analysis report (FSAR) used to establish the design basis for tipover events. Holtec did not fully evaluate the cumulative impact of the MOE changes or apply them consistently within the licensing basis. As a result, the NRC issued three Severity Level IV violations to Holtec for noncompliance with 10 CFR 72.48 requirements (see NRC Inspection Reports 07201014/2022-201, Holtec International (ML23145A175), and 07201014/2022-201, Holtec International, Inc. - Notice of Violation (ML24016A190)).
When the licensee (also referred to as a general licensee) chooses to adopt a change the CoC holder made pursuant to a CoC holder's change authority under 10 CFR 72.48 (referred to herein as a CoC holder-generated change), the licensee must perform a separate review using the requirements of 10 CFR 72.48(c). Accordingly, when the licensee chooses to adopt a CoC holder-generated change, and that change results in a non-conforming cask, there is a violation of 10 CFR 72.48 and certain provisions of 10 CFR 72.212 by the licensee, in addition to a CoC holder violation of 10 CFR 72.48.
In support of the 2022 loading campaign, the licensee adopted Holtecs generic design change, as documented in the "10 CFR 72.212 Report of Evaluations for HI-STORM FW System, Rev 5, and subsequently loaded casks using the CBS basket design. Because the CoC holder-generated change was found to be noncompliant by the NRC, the loaded casks at Watts Bar were also rendered non-conforming.
Corrective Actions: The licensee entered this into their corrective action program with actions to restore compliance with the 10 CFR 72.212 provisions that require each cask to conform to the terms, conditions, and specifications of a CoC or an amended CoC listed in 10 CFR 72.214.
Corrective Action References: 1907257
Enforcement:
Significance/Severity: The licensees failure to request that the CoC Holder obtain an amendment prior to implementing the change was determined to be of Severity Level IV significance based on the guidance in Section 1.2.6.D of the NRC's Enforcement Manual.
The severity of the violation was determined based on its very low safety significance, as documented in NRC memorandum titled Safety Determination of a Potential Structural Failure of the Fuel Basket During Accident Conditions for the HI-STORM 100 and HI-STORM Flood/Wind Dry Cask Storage Systems (ADAMS Accession No. ML24018A085) and its similarity with violation example 6.1.d.2 in the NRCs Enforcement Policy.
Violation: 10 CFR 72.48 (c)(1) requires in part that licensee or certificate holder may make changes in the facility or spent fuel storage cask design as described in the FSAR (as updated), without obtaining:
- (ii) CoC amendment submitted by the certificate holder pursuant to § 72.244 if:
- (c) The change, test, or experiment does not meet any of the criteria in paragraph (c)(2) of this section.
10 CFR 72.48(c)(2) requires in part that a general licensee shall request that the certificate holder obtain a CoC amendment, prior to implementing a proposed change, if the change would: (viii) Result in a departure from an MOE described in the FSAR used in establishing the design bases or in the safety analyses.
10 CFR 72.48(d)(1) requires in part that the licensee shall have a written evaluation which provides the bases for the determination that the change does not require a CoC amendment pursuant to 72.48(c)(2).
10 CFR 72.212(b)(3) requires in part that a general licensee must ensure that each cask used by the general licensee conforms to the terms, conditions, and specifications of a CoC or an amended CoC listed in 72.214.
Contrary to the above, since the 2022 loading campaign, the licensee failed to:
- (1) request Holtec, the certificate holder, obtain a CoC amendment for a change to the CBS cask design that resulted in a departure from an MOE described in the FSAR;
- (2) have a written evaluation providing the bases for the determination that the adopted change did not require a CoC amendment; and
- (3) ensure that the affected casks conformed to the terms, conditions, and specifications of the applicable CoC.
Specifically, Watts Bar's 10 CFR 72.48 titled 10 CFR 72.212 Report of Evaluations for HI-STORM FW System, Rev 5, failed to identify that the CBS variant design change resulted in a departure from a method of evaluation described in the FSAR used in establishing the design bases, failed to request the certificate holder obtain a CoC amendment pursuant to 10 CFR 72.244, and failed to ensure each cask conformed to the terms conditions, and specifications of a CoC or an amended CoC listed in 72.214, prior to using the CBS variant design.
Basis for Discretion: Section 9.4 of the Enforcement Policy, titled "Enforcement Discretion for General Licensee Adoption of Certificate of Compliance Holder-Generated Changes under 10 CFR 72.48" (ML25224A097), states that NRC will exercise enforcement discretion and not issue an enforcement action to a GL, for a noncompliance with the requirements of paragraphs (c)(1) and
- (2) and (d)(1) of 10 CFR 72.48 and with provisions of 10 CFR 72.212 that require GLs to ensure use of casks that conform to the terms, conditions and specifications of a CoC listed in 10 CFR 72.214, when the noncompliance results from a CoC holders failure to comply with 10 CFR 72.48 for a CoC holder-generated change. In support of the 2022 loading campaign, the licensee adopted a generic CoC holder design change (the CBS basket variant) and subsequently loaded the casks. On January 30, 2024, the NRC issued a notice of violation to the CoC holder, identifying the noncompliance, for the generic design change associated with the CBS basket variant (ML24016A190). As a result, the licensee became noncompliant due to the CoC holders failure to comply with 10 CFR 72.48 for the CoC holder-generated change. Since this violation meets the criteria of Section 9.4 of the policy, the NRC is exercising enforcement discretion by not issuing an enforcement action for this violation.
Failure to Demonstrate Effective Control of Performance of a Maintenance Rule System Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000390,05000391/2025003-01 Open/Closed
[P.1] -
Identification 71111.12 NRC inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants," paragraph (a)(2), for the licensee's failure to demonstrate effective control of performance of the Watts Bar unit 2 main steam valve vault room ventilation system through performance of appropriate preventive maintenance.
Description:
While reviewing reliability and availability data for the main steam valve vault ventilation system (system 030, function 030-Z) from July 10, 2023, to July 31, 2025, the inspectors identified 371 instances, as documented in station logs, of the temperature in the unit 2 main steam valve vault room exceeding its abnormal temperature limit of 145°F.
However, the vast majority of these instances were not accompanied by a condition report or documented assessment of the condition.
Technical instruction 0-TI-119, "Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting - 10CFR50.65," describes the performance criteria for the main steam valve vault room ventilation system as:
"The performance threshold is not more than one functional failure monitored over a 24-month interval for each valve vault per unit. Functional failure is defined as failure of the ventilation system to maintain acceptable temperatures in the Main Steam Valve Vaults. The determination of functional failure is based on the number of fans in service and the outside air temperature. This determination may require evaluation by Engineering on a case-by-case basis."
UFSAR Section 9.4.3.3.7, "Auxiliary Building Miscellaneous Ventilation and Air Conditioning System," states in part that the main steam valve vault ventilation exhaust airflow is regulated to maintain an adequate environment for the main steam safety valves.
Licensee Drawing 0-47E235-76, "Environmental Data Environment - Harsh EL 729.0,"
Table 5, defines the abnormal maximum temperature for the south main steam valve vault room as 145°F.
Licensee procedure NPG-SPP-03.4, Maintenance Rule, Performance Indicator Monitoring, Trending and Reporting - 10CFR50.65, Section 3.2.5A directs licensee staff to change monitoring status from (a)(2) to (a)(1) when an adverse trend in monitored data is occurring and causes are not understood or are not being effectively addressed such that exceeding any of the maintenance rule performance criteria may occur before resolution.
A review of the maintenance history for the main steam valve vault room exhaust fans revealed no preventive or corrective maintenance history. The inspectors determined that the demonstration of effective control of the performance of the system through preventive maintenance had not been technically justified due to the number of failures to maintain temperatures within environmental abnormal limits.
Corrective Actions: The licensee has documented the issue in the corrective action program and plans to present the issue to the maintenance rule expert panel for review.
Corrective Action References: CR 2028427
Performance Assessment:
Performance Deficiency: Failure of the licensee to demonstrate that the main steam valve vault room ventilation system performance was being effectively controlled through appropriate preventive maintenance, is a performance deficiency reasonably within the licensee's ability to foresee and prevent.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the number of failures to control temperature within the abnormal temperature limits and the lack of system maintenance indicated that the performance of the main steam valve vault room ventilation system was not effectively controlled through appropriate preventive maintenance, and the system was not moved to 10 CFR 50.65(a)(1) for monitoring. This is similar to IMC 0612 Appendix E, Example 8.g.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2, Mitigating Systems Screening Questions, Section A, the screening questions were all answered "No." Therefore, the inspectors determined the finding was of very low safety significance (Green).
Cross-Cutting Aspect: P.1 - Identification: The organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, a large number of instances of exceeding the abnormal temperature limit were not documented in the corrective action program or evaluated.
Enforcement:
Violation: 10 CFR 50.65(a)(1) states in part that each holder of an operating license for a nuclear power plant under this part shall monitor the performance or condition of SSCs, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSCs, as defined by 10 CFR 50.65(b), are capable of fulfilling their intended functions.
10 CFR 50.65(a)(2) states in part that monitoring, as specified in 10 CFR 50.65(a)(1), is not required where it has been demonstrated that the performance or condition of an SSC is being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remains capable of performing its intended function.
Contrary to the above, since July 10, 2023, the licensee failed to demonstrate that the performance or condition of the main steam valve vault room ventilation system was effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Specifically, as demonstrated through 371 examples of the failure to maintain temperature within abnormal environmental limits without subsequent evaluation, performance indicated that the system was not being effectively controlled through appropriate preventive maintenance, and the system was not moved to 10 CFR 50.65(a)(1) for performance monitoring.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Incorrect Rod Control Setup Resulted in Unanticipated Control Rod Withdrawal Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000391/2025003-02 Open/Closed
[H.11] -
Challenge the Unknown 71111.12 NRC inspectors identified a Green finding and associated non-cited violation of Watts Bar Unit 2 Technical Specification 5.7.1.1a, "Procedures," for the licensee's failure to ensure that appropriate instructions were established in a plant startup procedure associated with the control rod drive system. This directly resulted in improper setup of the control rod bank overlap circuitry for the unit 2 control rod banks and resulted in unanticipated and unplanned withdrawal of control rod bank B.
Description:
On July 16, 2025, NRC inspectors observed reactor startup following the unit 2 forced outage. During the start-up, withdrawal of control rod banks was briefed. Per the reactivity management plan and procedure 2-SOI-85.01, "Control Rod Drive Indication System," the control rod banks were to be withdrawn in sequence with proper overlap observed. During the withdrawal of control rod bank A, control rod bank B also withdrew.
Operators detected the unanticipated control rod withdrawal and reinserted the control rod banks in accordance with 2-SOI-85.01 and documented the issue in CR 2026418.
During subsequent troubleshooting under WO 125479975, it was discovered that the bank overlap controller indicated that power was available to withdraw control rod banks A and B, which was not the expected condition. Thumbwheel S1 was found at the 229 step position.
CR 2028204 documented the performance analysis for the unit 2 startup and determined that the operations crew had briefed and set the control rod drive logic cabinet thumbwheels S1 through S6 to 229 steps, instead of as required per nuclear operating book (NOB) sheet A7.
CR 2026422 documented a procedure change request to change 2-SOI-85.01, Section 5.3 (Closing Reactor Trip Breakers A & B), Step 16 from "ENSURE Logic Cabinet Thumbwheels, S1 thru S6, set per last performance of 2-TI-85.005, Quarterly Resetting of Full Out Rod Positions" to "ENSURE Logic Cabinet Thumbwheels, S1 thru S6, set as required by NOB Sheet A7." This procedure change was not in place prior to the unit start-up on July 16, 2025; the change to the procedure was implemented on August 7, 2025.
Corrective Actions: The condition was corrected through setting the S1 thumbwheel to the correct overlap setting and resetting the bank overlap circuitry; normal rod motion was observed. The issue was documented in the corrective action program.
Corrective Action References: CR 2028204
Performance Assessment:
Performance Deficiency: Failure to ensure that the procedure was adequate for the circumstances was a performance deficiency within the licensee's ability to foresee and prevent. Specifically, the control rod drive and indication system, system operating instruction 2-SOI-85.01, failed to contain the appropriate reference for setting the control bank overlap logic cabinet. This deficiency directly resulted in unanticipated control rod withdrawal.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the lack of reference to the NOB sheet resulted in a crew briefing and setting of the bank overlap logic to a condition that directly resulted in rod withdrawal that was neither anticipated nor per the approved reactivity management plan for the reactor startup. This directly challenged the reactivity control area of the configuration control attribute.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. All questions in Exhibit 3, Barrier Integrity Screening Questions, Section A, were answered "NO"; therefore, the inspectors determined the finding was of very low safety significance (Green).
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, personnel failed to challenge the condition of the bank overlap controller which was indicating that power was available to control banks A and B.
Enforcement:
Violation: Watts Bar Unit 2 Technical Specification 5.7.1.1a, "Procedures," requires in part that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in the revision of Regulatory Guide 1.33 as specified in the TVA Nuclear Quality Assurance Plan (TVA-NQA-PLN89-A).
TVA-NQA-PLN89-A, Revision 44, Section 9.9.2, "Plant Reviews," requires in part that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Regulatory Guide 1.33, Revision 2, Appendix A, Section 3, "Procedures for Startup, Operation, and Shutdown of Safety-Related PWR Systems," requires in part that instructions for energizing, filling, venting, draining, startup, shutdown, and changing modes of operation should be prepared, as appropriate, for the listed systems including the control rod drive system.
Contrary to the above, from October 22, 2015 (initial plant start-up), through August 7, 2025, the licensee failed to establish appropriate instructions in a startup procedure associated with the control rod drive system. Specifically, procedure 2-SOI-85.01 did not contain adequate steps to ensure that control rod bank overlap was set correctly, which directly resulted in unanticipated control rod withdrawal during a unit 2 reactor startup on July 16, 2025.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Degraded Public Address System 71152A The inspectors reviewed condition reports 2023435 and 2021575 for an in-depth review of the degraded public address system at Watts Bar. The inspection conclusions associated with this review are documented in inspection report 05000390, 05000391/2025090 in Inspection Results Section 71152A.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On December 8, 2025, the inspectors presented the integrated inspection results to Chris Reneau, Site Vice President, and other members of the licensee staff.
- On November 20, 2025, the inspectors presented the ISFSI CBS basket inspection results to Brian Cupp, Dry Cask Storage Manager, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
CR 2024874
During WBN 2025 QA Audit unannounced fire drill the
following drill evaluation identified critical items not met.
07/08/2025
Corrective Action
Documents
Resulting from
Inspection
CR 2027794
NRC Identified Door A039 (SIP 2A-A Room) has loose
gasket material
07/22/2025
AUX-0-676-01
Pre-fire plan Auxiliary Building 676.0' Elevation
003
AUX-0-692-01
Pre-fire plan Auxiliary Building 692.0' Elevation
003
AUX-0-692-03
Pre-fire plan Auxiliary Building 692.0' Elevation
008
AUX-0-713-01
Pre-fire plan Auxiliary Building 713.0' Elevation
2
AUX-0-713-03
Pre-fire plan Auxiliary Building 713.0' Elevation
005
AUX-0-772-01
Pre-fire plan Auxiliary Building 772.0' Elevation
Fire Plans
AUX-0-772-02
Pre-fire plan Auxiliary Building 772.0' Elevation
005
Procedures
0-TI-12.16
Diesel Generator Outage T/S or SR Contingency Actions
Revision 18
Corrective Action
Documents
CR 20255847
Unit 2 Reactor Trip
07/13/2025