IR 05000259/2025003
| ML25349B313 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/18/2025 |
| From: | Renee Taylor NRC/RGN-II/DORS/PB5 |
| To: | Erb D Tennessee Valley Authority |
| References | |
| EA-NMSS-2023-0002, EAF-NMSS-2025-0216 IR 2025003, IR 2024001 | |
| Download: ML25349B313 (0) | |
Text
SUBJECT:
BROWNS FERRY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000259/2025003, 05000260/2025003, 05000296/2025003, AND 07200052/2024001 AND EXERCISE OF ENFORCEMENT DISCRETION
Dear Delson Erb:
On September 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant. On November 17, 2025, the NRC inspectors discussed the results of this inspection with Daniel Komm, 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 consistent with Section 2.3.2 of the Enforcement Policy.
December 18, 2025 The NRC also 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 (MPC) 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 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 Browns Ferry 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 Browns Ferry 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. 05000259, 05000260, 05000296, and 07200052 License Nos. DPR-33, DPR-52, and DPR-68
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000259, 05000260, 05000296 and 07200052
License Numbers:
Report Numbers:
05000259/2025003, 05000260/2025003, 05000296/2025003, and
07200052/2024001
Enterprise Identifier:
I2025-003-0025 and I2024-001-0133
Licensee:
Tennessee Valley Authority
Facility:
Browns Ferry Nuclear Plant
Location:
Athens, Alabama
Inspection Dates:
July 1, 2025 to September 30, 2025
Inspectors:
S. Billups, Resident Inspector
P. Cooper, Senior Reactor Inspector
A. Craig, Project Engineer
D. Neal, Health Physicist
K. Pfeil, Resident Inspector
T. Steadham, 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 Browns Ferry 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 Monitor Standby Monitoring Parameters Leads to Unit 3 Reactor Core Isolation Cooling Inoperable Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000296/2025003-01 Open/Closed
[H.12] - Avoid Complacency 71111.15 A self-revealed Green finding and associated non-cited violation (NCV) of Browns Ferry Nuclear Unit 3 Technical Specifications (TS) Limiting Condition for Operations (LCO) 3.5.3 and 3.0.4 was identified when the licensee failed to comply with procedure NPG-SPP-09.0.1,
Conduct of System Engineering and Equipment Reliability, Revision 14. Specifically, the licensee failed to monitor the reactor core isolation cooling (RCIC) system as required by the applicable system monitoring plan. As a consequence, a component failure that rendered the system inoperable could have been identified and corrected prior to operating in a condition prohibited by TS.
Failure to Perform a Seismic Evaluation of a Degraded Safety-Related Check Valve Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000296/2025003-02 Open/Closed
[H.12] - Avoid Complacency 71152A Inspectors identified a Green finding and associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, when the licensee failed to follow quality related procedure OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking. Specifically, following the identification of a degraded check valve in the right bank starting air system for the 3B standby diesel generator, the licensee failed to consider the effects of a seismic event on the non-seismic components upstream of this check valve. As a consequence, the licensee did not have a complete basis for operability when the redundant left bank starting air system was removed from service for planned maintenance.
Additional Tracking Items
Type Issue Number Title Report Section Status EDG EAF-NMSS-2025-0216 Interim Enforcement Policy (IEP) Associated with the Continuous Basket Shim 60855 Closed
PLANT STATUS
Unit 1 began the inspection period at 100 percent rated thermal power (RTP). On August 2, 2025, a malfunction in the main feedwater control system caused an automatic scram.
Following completion of the repairs, operators restarted the unit on August 3, 2025. On August 7, 2025, the unit was returned to 100 percent RTP. On August 8, 2025, operators lowered reactor power to 65 percent RTP for a control rod pattern adjustment. On August 9, 2025, the unit was returned to 100 percent RTP. On September 5, 2025, operators lowered reactor power to 56 percent RTP for a control rod sequence exchange. On September 7, 2025, the unit was returned to 100 percent RTP, where it operated for the remainder of the inspection period.
Unit 2 began the inspection period at 100 percent RTP. On August 30, 2025, operators lowered reactor power to 62 percent RTP for a control rod sequence exchange. On August 31, 2025, the unit was returned to 100 percent RTP where it operated for the remainder of the inspection period.
Unit 3 began the inspection period at 100 percent RTP. On August 4, 2025, operators manually scrammed the reactor to perform planned repairs to address a mechanical seal leak on the 3B reactor recirculation pump. Following completion of the repairs, operators restarted the unit on August 9, 2025. On August 14, 2025, the unit was returned to 100 percent RTP. On September 26, 2025, operators lowered reactor power to 65 percent RTP for a control rod sequence exchange. On September 27, 2025, the unit was returned to 100 percent RTP, where it operated 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
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (1 Sample)
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 3, primary containment on August 8, 2025
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated a licensed operator requalification session in the unit 3 simulator which included emergency equipment cooling water pump trip, an earthquake, all four emergency diesel generators trouble alarms due to air leak, all four emergency diesel generators failure to start, loss of off-site power leading to a station blackout, and an automatic scram on August 26, 2025. This training session required the crew to enter various abnormal operating instructions, emergency operating instructions, and emergency plan implementing procedures to control the plant and appropriately classify the emergency.
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (7 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)maintenance risk assessment during unit common standby gas treatment "C" maintenance on July 16, 2025 (2)emergent work control for unit 3 west scram discharge volume high level switch during a conservative operations alert on July 25, 2025 (3)maintenance risk assessment during unit 3 diesel generator 3A planned maintenance outage during conservative operation alert and power supply alert on August 6, 2025 (4)emergent work control for unit 3 standby liquid control degraded test line throttle valve on August 20, 2025 (5)emergent work on unit 3 reactor core isolation cooling system failed ramp generator signal converter while performing flow rate test on August 25, 2025 (6)maintenance risk assessment during unit common control room emergency ventilation "A" planned maintenance on September 3, 2025 (7)maintenance risk assessment during main battery bank 1 planned maintenance and testing outage on September 26, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1)condition report (CR) 2020461, unit 3 standby diesel generator 3C left bank air compressor discharge check failed to seat on July 15, 2025
- (2) CR 2032214, unit 3 standby liquid control test line throttle valve stem to disc separation on August 22, 2025 (3)evaluation of seismic interactions with scaffold 25-595-1 and unit 1 loop II residual heat removal drywell spray piping on August 29, 2025
- (4) CR 2033773, unit 3 reactor core isolation cooling system failed ramp generator signal converter on September 24, 2025
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated the following:
(1)unit 1 forced outage activities (August 1, 2025, through August 4, 2025)
(2)unit 3 forced outage activities (August 4, 2025, through August 9, 2025)
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) (4 Samples)
(1)work order (WO) 124553326, unit 2 residual heat removal pump "B" suppression pool suction valve in-service testing after actuator maintenance on July 14, 2025
- (2) WO 124528313, unit 3 3A standby diesel generator surveillance testing after planned maintenance and component replacement on August 11, 2025
- (3) WO 124552633, unit 3 reactor core isolation cooling system rated flow surveillance test after ramp generator signal converter replacement on September 9, 2025
- (4) WO 125190063, perform leak rate test on unit 2 residual heat removal heat exchanger relief valve after replacement on September 25, 2025
Surveillance Testing (IP Section 03.01) (1 Sample)
- (1) WO 124444430, unit common control room emergency ventilation "B" iodine removal efficiency surveillance test on September 4, 2025
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) WO 124577230, unit 1 residual heat removal loop II rated flow test on August 27, 2025
Reactor Coolant System Leakage Detection Testing (IP Section 03.01) (1 Sample)
(1)unit 2 unidentified drywell leakage monitoring due to increase leak rate on July 1, 2025
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
- (1) The inspectors observed and evaluated an emergency preparedness drill on August 13, 2025. Events included the loss of raw cooling water pump 2A, reactor water cleanup pump 2B, and fuel pool cooling pump 2A due to localized seismic activity (earthquake), a scram without control rod insertion, loss of 4KV shutdown board "B," feedwater line break resulting in low reactor water level and fuel damage, and entry into the severe accident mitigation guidelines.
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)===
(1)unit 1 (July 1, 2024, through June 30, 2025)
(2)unit 2 (July 1, 2024, through June 30, 2025)
(3)unit 3 (July 1, 2024, through June 30, 2025)
MS09: Residual Heat Removal Systems (IP Section 02.08) (3 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
(2)unit 2 (July 1, 2024, through June 30, 2025) (3)unit 3 (July 1, 2024, through June 30, 2025)
MS10: Cooling Water Support Systems (IP Section 02.09) (3 Samples)
(1)unit 1 (July 1, 2024, through June 30, 2025)
(2)unit 2 (July 1, 2024, through June 30, 2025) (3)unit 3 (July 1, 2024, through June 30, 2025)
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) CR 2007162, corrective actions to address repetitive failures of the unit 3 standby diesel generator 3B right bank starting air check valve on August 27, 2025
- (2) CR 1997594, unit 2 residual heat removal relief valve failed lift test acceptance criteria on September 25,
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) 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-2025-0216: 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 multipurpose canister (MPC) fuel basket, known as the CBS 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 tip-over 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 Browns Ferry 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: 1907246
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, 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, Browns Ferry'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 Monitor Standby Monitoring Parameters Leads to Unit 3 Reactor Core Isolation Cooling Inoperable Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000296/2025003-01 Open/Closed
[H.12] - Avoid Complacency 71111.15 A self-revealed Green finding and associated NCV of Browns Ferry Nuclear Unit 3 TS LCO 3.5.3 and 3.0.4 was identified when the licensee failed to comply with procedure NPG-SPP-09.0.1, Conduct of System Engineering and Equipment Reliability, Revision 14. Specifically, the licensee failed to monitor the RCIC system as required by the applicable system monitoring plan. As a consequence, a component failure that rendered the system inoperable could have been identified and corrected prior to operating in a condition prohibited by TS.
Description:
The RCIC turbine control system relies on a series of signal pathways to regulate valve position and ensure proper system response. Under normal operation, the ramp generator signal converter (RGSC) provides a reference signal to the electric governor-magnetic pickup (EG-M) control box. The EG-M uses this reference to generate an output signal that ultimately positions the turbine control valve to match the flow demand set by the RCIC flow controller. In this way, the RGSC serves as the starting point for the control chain that ultimately regulates steam flow to the RCIC turbine and enables the pump to deliver the required flow.
On August 19, 2025, the Unit 3 RCIC system failed its quarterly surveillance test because the turbine control valve operated erratically, preventing the system from achieving stable flow.
Troubleshooting identified the cause as a failed low output signal from the RGSC. In this failed state, the RGSC drove the EG-M into a negative voltage range that caused the control system to issue a valve-closed demand regardless of the actual demand from the RCIC controller. The licensee replaced the failed RGSC which restored the control system to functional. The licensee successfully reperformed the surveillance and declared RCIC operable on August 22, 2025.
A later review of historical data revealed that the RGSC failed almost a month earlier. On July 20, 2025, at 11:54 a.m., the monitored computer point (ANA022) for the EG-M shifted from a normal value of -0.45 V to an abnormally low -8.21 V. This abnormal signal was consistent with a closed-valve demand from the RGSC, directly prevented the turbine control valve from stabilizing during the August 19 test, and rendered the RCIC system inoperable until the fault was corrected.
NPG-SPP-09.0.1, Revision 14, Section 3.2.1, System Classification, states that for Tier 1 systems, such as RCIC, a system monitoring plan will be developed and monitoring/trending will be performed. The RCIC system monitoring plan included the requirement to monitor computer point ANA022 weekly but because of a human performance error, the system engineer was not monitoring that point. If the licensee had monitored that point as per the monitoring plan, the licensee would have identified the failed RGSC within a week. This would have allowed sufficient time for troubleshooting and repairs before exceeding the TS LCO allowed outage time of 14 days.
During this time period that RCIC was inoperable, unit 3 entered and exited an outage.
Because the TS generally prohibits entry into a mode of applicability with required equipment inoperable, the inspectors reviewed the time periods that the RGSC was in a failed state but RCIC was not required to be operable. Specifically, the reactor remained in mode 1 until August 4, 2025, when the reactor entered mode 3 to begin the planned maintenance outage. Following a normal reactor cooldown, reactor steam dome pressure fell to less than 150 psig at 1:32 a.m. on August 5, 2025. Following repairs, the reactor entered mode 2 and exceeded 150 psig reactor steam dome pressure on August 9, 2025.
Corrective Actions: The licensee replaced the RGSC and successfully reperformed the RCIC surveillance. The licensee reviewed system monitoring plans to ensure all systems were monitored in accordance with their approved plans to ensure there were no additional similar gaps.
Corrective Action References: CR 2033486, CR 2034237, WO 125556690, WO 125556569, WO 125559901, WO 124552633
Performance Assessment:
Performance Deficiency: The failure to comply with procedure NPG-SPP-09.0.1, Revision 14, was a performance deficiency. Specifically, the licensee failed to monitor all required system parameters as required by Reactor Core Isolation Cooling System Monitoring Plan dated September 28, 2024.
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 failure to monitor the RCIC standby monitoring parameters directly contributed to RCIC being inoperable for greater than the allowed outage time.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Because RCIC is a single train system and the PRA function was lost for a period greater than the TS allowed outage time, the issue screens to a Detailed Risk Evaluation (DRE). A regional Senior Reactor Analyst (SRA) performed a DRE using SAPHIRE 8, Version 8.2.11 and the Browns Ferry Unit 3 SPAR Model Version 8.82, dated August 14, 2023. Since the date of the failure can be established using plant computer records, the SRA considered exposure time to be T + repair time while the until was in a mode where RCIC was required to be operable. Thus, exposure time was from July 20, 2025, until August 5, 2025, when unit 3 entered a maintenance outage, and from August 9, 2025, until August 22, 2025, when RCIC was returned to an operable status. This is a period of 29 days. The SRA modeled the condition by setting RCI-TDP-FS-TRAIN, RCIC PUMP FAILS TO START, to TRUE. The dominant accident sequence was a loss of condenser heat sink, with a failure of high-pressure coolant injection to run, and operators failing to depressurize and failing to align the emergency high pressure makeup system. The change in core damage frequency and large early release frequency were both less than 1E-7, corresponding to a finding of very low safety significance (GREEN).
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, the licensee failed to implement the monitoring plan due to the incorrect assumption that the weekly standby parameter monitoring was not necessary.
Enforcement:
Violation: Browns Ferry Nuclear Unit 3 TS LCO 3.5.3, requires, in part, that the RCIC system shall be operable while in either Mode 1 or Modes 2 and 3 with reactor steam dome pressure greater than 150 psig. If the RCIC system is inoperable in an applicable mode, LCO 3.5.3, Condition A, requires, in part, restoring the system to operable status within 14 days. If the required actions for Condition A are not met within the established completion time, Condition B requires the unit to be in mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to reduce reactor steam dome pressure to less than or equal to 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Contrary to the above, from August 4, 2025, to August 5, 2025, the licensee failed to place unit 3 in mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the RCIC system was inoperable for 14 days while in either Mode 1 or Modes 2 and 3 with reactor steam dome pressure greater than 150 psig. Specifically, RCIC was inoperable from July 20, 2025, with the unit in mode 1 and remained inoperable in a mode of applicability until August 5, 2025, when the unit was in mode 3 and steam dome pressure fell below 150 psig.
Browns Ferry Nuclear Plant, Unit 3 TS LCO 3.0.4, requires, in part, that when an LCO is not met, entry into a mode or other specified condition in the applicability shall only be made when the associated actions to be entered permit continued operation in the mode or other specified condition in the applicability for an unlimited period of time.
Contrary to the above, on August 9, 2025, the licensee entered a mode or other specified condition in the applicability when the associated actions to be entered did not permit continued operation in the mode or other specified condition in the applicability for an unlimited period of time. Specifically, Unit 3 RCIC was inoperable when, while starting up the unit from a maintenance outage, the unit was in mode 2 and steam dome pressure rose above 150 psig on August 9, 2025.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Perform a Seismic Evaluation of a Degraded Safety-Related Check Valve Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000296/2025003-02 Open/Closed
[H.12] - Avoid Complacency 71152A Inspectors identified a Green finding and associated NCV of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, when the licensee failed to follow quality related procedure OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking. Specifically, following the identification of a degraded check valve in the right bank starting air system for the 3B standby diesel generator, the licensee failed to consider the effects of a seismic event on the non-seismic components upstream of this check valve. As a consequence, the licensee did not have a complete basis for operability when the redundant left bank starting air system was removed from service for planned maintenance.
Description:
Each standby diesel generator at Browns Ferry Nuclear Plant is equipped with two fully redundant air start systemsdesignated as the left bank and right bankeither of which is independently capable of starting the engine. In accordance with TS 3.8.3, Condition E, diesel generator operability requires only one functional air start subsystem, provided that the associated starting air receiver pressure remains greater than 165 psig. Consequently, if both air start subsystems are rendered inoperable, the associated diesel generator must be declared inoperable.
On April 6, 2025, during performance of surveillance test procedure SR-3.8.1.1(3B), Diesel Generator 3B Monthly Operability Test, the right bank starting air compressor discharge check valve (3-CKV-086-0500B) failed to properly seat following compressor shutdown, as required by step 6.5.2[7]. This condition was documented in CR 2004275; however, no operability determination was recorded. Subsequent review of planned maintenance activities revealed that the licensee intended to remove the left bank from service, which would leave the degraded right bank as the sole air start system. Inspectors raised concerns regarding the seismic qualification of upstream componentsspecifically the air dryer and air compressorassociated with the right bank. These components were confirmed to be non-seismically qualified for pressure retention, meaning they are not designed to maintain structural integrity or pressure containment following a design-basis seismic event. Therefore, if a design-basis seismic event were to occur, failure of these components could result in depressurization of the right bank air receivers, thereby compromising the ability to start the 3B standby diesel generator. In response to these concerns, the licensee deferred left bank maintenance until the affected check valve was replaced. The replacement of 3-CKV-086-0500B was completed on April 29, 2025.
Historical review indicated that 3-CKV-086-0500B had previously been replaced on May 5, 2022, with the first post-replacement failure to seat occurring on May 29, 2022. Between May 2022, and April 2025, the valve exhibited intermittent failures to seat during SR 3.8.1.1(3B)testing, with at least eight documented instances of nonconformance. During this timeframe, the left bank air start system was periodically removed from service for maintenance, with the longest outage lasting fewer than five days. Each instance of left bank unavailability resulted in reliance on the right bank air start system to maintain diesel generator operability under all credible design basis conditions. However, the right banks continued operability was contingent on the assumption that its air receiver pressure would remain above the TS-required threshold, without consideration of potential depressurization due to seismic-induced failure of non-qualified upstream components.
Review of prior corrective actions revealed that the licensee consistently concluded that the failure of 3-CKV-086-0500B to seat did not constitute an operability concern, based solely on the observation that right bank air receiver pressure remained above 165 psig. However, no documented evaluation addressed the ability of the right bank air receiver to maintain required pressure following a design basis seismic event. The inspectors determined that procedure OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking, was applicable to these assessments. Revisions 29 through 33 of OPDP-8 explicitly requires that operability determinations be based on the capability of structures, systems, and components to perform their safety functions under specific design basis eventsnot merely under current operating conditions. Accordingly, the inspectors concluded that the operability determinations associated with the degraded condition of 3-CKV-086-0500B failed to meet the procedural requirements of OPDP-8.
Corrective Actions: WO 125281021 refurbished 3-CKV-086-0500B to restore functionality and surveillance test SR-3.8.1.1(3B) was satisfactorily reperformed to restore operability.
Corrective Action References: CRs 1779634, 1799069, 1813331, 1813643, 1814485, 1866096, 1884223, 1970047, 1997586, 2004275, 2007162
Performance Assessment:
Performance Deficiency: The licensees failure to follow procedure OPDP-8 was a performance deficiency. Specifically, the licensee failed to ensure the operability decision considered the safety analyses of design basis events. Not fully evaluating a known deficiency of the 3B standby diesel generator right bank starting air system check valve and executing planned maintenance on the opposing left bank starting air system without fully understanding the operability status of 3B diesel generator was reasonably within the licensees 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, this performance deficiency is similar to Example 3.k in IMC 0612, Appendix E, in that the inspectors identified the licensee had not addressed the seismic effects on the starting air system such that the operability conclusion was truly challenged, and the licensee had to perform actions (e.g. a check-valve leakage test) to demonstrate operability.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the performance deficiency using Exhibit 2 of Appendix A, Mitigating Systems Screening Questions, and determined that the issue screened to Green based on answering No to all screening questions.
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, the licensee failed to consider a seismic event in their evaluation of operability for the 3B diesel generator with the degradation of the right bank starting air check valve. This complacency was present with the eight identified failures of the monthly operability test for the 3B diesel generator from May of 2022 until April of 2025.
Enforcement:
Violation: 10 CFR 50, Appendix B, Criterion V, states, in part, that Activities affecting quality shall be prescribed by documented procedures and shall be accomplished in accordance with those procedures. OPDP-8, Revisions 29 through 33, Section 3.3.4.F.5, states, in part, the structures, systems, and components operability requirements are based on safety analyses of specific design basis events for one operating mode or specified condition of operations.
Contrary to the above, from May 29, 2022, until April 29, 2025, the licensee failed to accomplish activities affecting quality in accordance with documented procedures.
Specifically, on multiple occasions since the identification of the initial leak-by on 3-CKV-086-0500B in May of 2022, the licensee did not consider a design basis seismic event when evaluating the operability of the 3B diesel generator right bank starting air system.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On November 17, 2025, the inspectors presented the integrated inspection results to Daniel Komm, 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
Miscellaneous
BFN-ODM-4.18
Protected Equipment
0036
Corrective Action
Documents
CR 2020461,
2032214,
2033486,
2033773,
2034237
Miscellaneous
Engineering evaluation for scaffold tag no. 25-595-1
08/28/2025
Operability
Evaluations
Past operability evaluation CR 2033486
08/28/2025
Work Orders
24444590,
24387222,
25556690,
25559901,
25556569
Work Orders
24553326,
24553277,
24444520,
24528313,
24662501,
24577230,
24444430,
2967421,
21351981,
25190063,
24552633,
25556569,
25556690,
25559901
71151
Miscellaneous
MSPI Derivation
Reports
MSPI Derivation Reports for Unreliability Index (URI) &
Unavailability Index (UAI)
71151
Miscellaneous
Unit 1 MSPI
U1 MSPI Margin Report_MS06, 09, 10
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Margin Report
71151
Miscellaneous
Unit 2 MSPI
Margin Report
U2 MSPI Margin Report_MS06, 09, 10
71151
Miscellaneous
Unit 3 MSPI
Margin Report
U3 MSPI Margin Report_MS06, 09, 10
Corrective Action
Documents
2007162,
1779634,
1799069,
1813331,
1813643,
1866096,
1884223,
1997586,
2004275,
1814485,
1970047,
2011280,
2004275,
1997594,
1997744,
1997745
Work Orders
2102380,
25281021,
23278111,
23626978,
2895304,
2884648,
2884649,
2884666,
2884667,
24467612,
24467613,
25190063,
25190064,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
23639889