IR 05000259/2025004
| ML26033A309 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/09/2026 |
| From: | Renee Taylor NRC/RGN-II/DORS/PB5 |
| To: | Erb D Tennessee Valley Authority |
| References | |
| IR 2025004 | |
| Download: ML26033A309 (0) | |
Text
SUBJECT:
BROWNS FERRY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000259/2025004, 05000260/2025004, AND 05000296/2025004
Dear Delson Erb:
On December 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant. On January 8, 2026, 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.
Two findings of very low safety significance (Green) are documented in this report. One of these findings involved a violation of NRC requirements. In addition, one Severity Level IV violation without an associated finding and one licensee-identified violation which was determined to be of very low safety significance (Green) are documented in this report. We are treating these violations as non-cited violations consistent with Section 2.3.2 of the Enforcement Policy.
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 finding not associated with a regulatory requirement 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.
February 9, 2026 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, and 05000296 License Nos. DPR-33, DPR-52, and DPR-68
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000259, 05000260, and 05000296
License Numbers:
Report Numbers:
05000259/2025004, 05000260/2025004, and 05000296/2025004
Enterprise Identifier:
I-2025-004-0025
Licensee:
Tennessee Valley Authority
Facility:
Browns Ferry Nuclear Plant
Location:
Athens, Alabama
Inspection Dates:
October 1 to December 31, 2025
Inspectors:
J. Alamudun, Reactor Inspector
S. Billups, Resident Inspector
D. Lanyi, Senior Operations Engineer
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. A licensee-identified non-cited violation is documented in report section: 7115
List of Findings and Violations
Failure to Perform Preventive Maintenance Leads to Dual Reactor Recirculation Pump Trips Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000296/2025004-01 Open/Closed None (NPP)71153 A self-revealed Green finding was identified when the licensee failed to classify the auxiliary power distribution contactor for the Unit 2 reactor recirculation pump variable frequency drives as a critical component in accordance with licensee procedure NEDP-12, Equipment Failure Trending, Revision 3. Specifically, the licensee misclassified the contactor as non-critical and this misclassification resulted in no PMs being performed on the component until its age-related failure caused a reactor trip. Had the licensee classified the contactor as critical, PMs would have been procedurally developed and performed to identify and correct degradation of the contactor prior to its failure.
Pressure Boundary Leaks on Reactor Recirculation Small Bore Piping Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000260/2025004-02 Open/Closed None (NPP)71153 A self-revealed Green non-cited violation (NCV) of Technical Specification (TS) 3.4.4,
"Reactor Coolant System Operational Leakage," was identified when the licensee failed to accurately model the cyclic forces for the small bore piping associated with the Unit 3 reactor recirculation system. As a result, a pressure boundary leak developed on two small bore reactor recirculation lines when flow-induced vibrations caused cyclic fatigue failure of both lines.
Main Steam Relief Valves Lift Settings Outside of Technical Specification Required Setpoints Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000260/2025004-03 Open/Closed Not Applicable 71153 A self-revealed Severity Level IV NCV of TS 3.4.3, "Safety/Relief Valves (S/RVs)," was identified when the licensee discovered, through as-found test results, that 4 of 13 main steam relief valves (MSRVs) had as-found lift settings outside of the +/-3 percent setpoint band required for their operability.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000296/2025-002-00 Unit 3, Reactor Core Isolation Cooling Inoperable Due to Failed Ramp Generating Signal Converter 71153 Closed LER 05000260/2025-001-00 Unit 2, Reactor SCRAM due to Low Electro-Hydraulic Control Pressure 71153 Closed LER 05000260/2025-001-01 Unit 2, Reactor SCRAM due to Low Electro-Hydraulic Control Pressure 71153 Closed LER 05000260/2025-002-00 Unit 2, Residual Heat Removal Suppression Pool Spray Inoperable Due to Missing U-Bolt 71153 Closed LER 05000260/2025-003-00 Unit 2, Primary Containment Isolation and Manual Reactor Trip Due to Dual Recirculation 71153 Closed LER 05000260/2025-003-01 Unit 2, Primary Containment Isolation and Manual Reactor Trip Due to Dual Recirculation Pump Trips 71153 Closed LER 05000260/2025-004-00 Unit 2, Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints 71153 Closed LER 05000296/2024-004-00 Unit 3, Reactor Coolant System Pressure Boundary Leakage Identified from Recirculation System Small Bore Piping 71153 Closed LER 05000296/2024-004-01 Unit 3, Reactor Coolant System Pressure Boundary Leakage Identified from Recirculation System Small Bore Piping 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at full (100 percent) rated thermal power (RTP). On December 5, 2025, operators reduced reactor power to 70 percent RTP to perform a control rod sequence exchange and turbine valve testing. The unit was returned to 100 percent RTP on December 6, 2025, where it remained for the remainder of the inspection period.
Unit 2 began the inspection period at full RTP. On December 12, 2025, operators reduced reactor power to 66 percent RTP to perform a control rod sequence exchange and turbine valve testing. The unit was returned to 100 percent RTP on December 13, 2025, where it remained for the remainder of the inspection period.
Unit 3 began the inspection period at full RTP. On November 21, 2025, operators reduced reactor power to 69 percent RTP to perform a control rod sequence exchange and turbine valve testing. The unit was returned to 100 percent RTP on November 22, 2025. On December 7, 2025, reactor power was reduced to 73 percent RTP for a control rod pattern adjustment. The unit was returned to 100 percent RTP on December 8, 2025, where it remained 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
71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance
Requalification Examination Results (IP Section 03.03) (1 Sample)
The licensee completed the annual requalification operating examinations and biennial written examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), "Requalification Requirements," of the NRC's "Operator's Licenses." The inspector performed an in-office review of the overall pass/fail results of the individual operating examinations, the crew simulator operating examinations, and the biennial written examinations in accordance with IP 71111.11, "Licensed Operator Requalification Program and Licensed Operator Performance." These results were compared to the thresholds established in Section 3.03, "Requalification Examination Results," of IP 71111.11.
- (1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam administered in November 2025 and the biennial written examinations completed in December 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 the common accident signal logic testing on October 30, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated a licensed operator requalification session in the simulator which included a main condensate pump trip, a main steam relief valve spurious opening, a tornado striking the plant, all four diesel generators failure to start, loss of off-site power leading to a station blackout, and an automatic scram on October 7, 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.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (5 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) Units 1, 2, and 3 emergency high pressure makeup crosstie function 7-D return to 10CFR50.65(a)(2) status on October 9, 2025
- (2) Units 1, 2, and 3 reactor recirculation system variable frequency drive equipment issue trends on October 17, 2025
- (3) Units 1, 2, and 3 4kV shutdown board and buses condition monitoring on November 6, 2025
- (4) Units 1, 2, and 3 function 575-B, unit loads, condition monitoring on November 6, 2025
- (5) Units 1, 2, and 3 CR105/109 motor starter contactor failures in multiple systems on November 7, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1)condition report (CR) 2040968, Unit 2 reactor core isolation cooling electrical junction box high surface temperature on October 1, 2025
- (2) CR 2041571, Unit 3 standby diesel generator "3B" lubricating oil immersion heater control temperature switch found out of tolerance on October 3, 2025
- (3) CR 2044150, Unit 1 core spray pump discharge vent line pressure boundary weld evaluation on October 23, 2025
- (4) CR 2051196, Unit 2 residual heat removal shutdown cooling suction header high pressure indication on December 15, 2025
- (5) CR 2054399, Unit 1 reactor core isolation cooling suppression pool inboard suction valve did not open on December 16, 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) (1 Sample)
(1)work order (WO) 124772979, Unit 2 residual heat removal heat exchanger "C" service water outlet valve test after spring pack replacement on October 6, 2025
Surveillance Testing (IP Section 03.01) (1 Sample)
- (1) WO 124528340, Unit 3 standby diesel generator 3B monthly operability test on October 8, 2025
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) (3 Samples)
- (1) Unit 1 (October 1, 2024, through September 30, 2025)
- (2) Unit 2 (October 1, 2024, through September 30, 2025)
- (3) Unit 3 (October 1, 2024, through September 30, 2025)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (3 Samples)
- (1) Unit 1 (October 1, 2024, through September 30, 2025)
- (2) Unit 2 (October 1, 2024, through September 30, 2025)
- (3) Unit 3 (October 1, 2024, through September 30, 2025)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (3 Samples)
- (1) Unit 1 (October 1, 2024, through September 30, 2025)
- (2) Unit 2 (October 1, 2024, through September 30, 2025)
- (3) Unit 3 (October 1, 2024, through September 30, 2025)
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02) (1 Sample)
- (1) The inspectors reviewed the licensees corrective action program to identify potential trends in public address (PA) speaker performance that might be indicative of a more significant safety issue.
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02) (9 Samples)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 50-260/2025-001-00, Unit 2, Reactor SCRAM due to Low Electro-Hydraulic Control Pressure (ADAMS Accession No. ML25191A210). The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensee's ability to foresee and correct and therefore was not reasonably preventable. No performance deficiency or violation of NRC requirements was identified. This LER is Closed.
- (2) LER 50-260/2025-001-01, Unit 2, Reactor SCRAM due to Low Electro-Hydraulic Control Pressure (ADAMS Accession No. ML25252A080). This LER is a revision to an LER that was previously reviewed in this inspection report. The inspectors reviewed the updated LER submittal. No performance deficiency or violation of NRC requirements was identified as a result of this review. This LER is Closed.
- (3) LER 50-260/2025-002-00, Unit 2, Residual Heat Removal Suppression Pool Spray Inoperable due to Missing U-bolt (ADAMS Accession No. ML25195A123). The inspectors determined that this LER was related to a licensee-identified non-cited violation that is documented in this report. No additional performance deficiency or violation of NRC requirements was identified as a result of this review. This LER is Closed.
- (4) LER 50-260/2025-003-00, Unit 2, Primary Containment Isolation and Manual Reactor Trip Due to Dual Recirculation Pump Trips (ADAMS Accession No. ML25210A472).
The inspectors determined that this LER was related to a Green finding that is documented in this report. No additional performance deficiency or violation of NRC requirements was identified as a result of this review. This LER is Closed.
- (5) LER 50-260/2025-003-01, Unit 2, Primary Containment Isolation and Manual Reactor Trip Due to Dual Recirculation Pump Trips (ADAMS Accession No.
ML25300A115). This LER is a revision to an LER that was previously reviewed in this inspection report. The inspectors reviewed the updated LER submittal. No additional performance deficiency or violation of NRC requirements was identified as a result of this review. This LER is Closed.
- (6) LER 50-260/2025-004-00, Unit 2, Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints (ADAMS Accession No. ML25237A236).
The inspectors determined that this LER was related to a Severity Level IV non-cited violation that is documented in this report. No additional performance deficiency or violation of NRC requirements was identified as a result of this review. This LER is Closed.
- (7) LER 50-296/2024-004-00, Unit 3, Reactor Coolant System Pressure Boundary Leakage Identified from Recirculation System Small Bore Piping (ADAMS Accession No. ML24339B794). The inspectors determined that this LER was related to a Green finding that is documented in this report. No additional performance deficiency or violation of NRC requirements was identified as a result of this review. This LER is Closed.
- (8) LER 50-296/2024-004-01, Unit 3, Reactor Coolant System Pressure Boundary Leakage Identified from Recirculation System Small Bore Piping (ADAMS Accession No. ML25114A051). This LER is a revision to an LER that was previously reviewed in this inspection report. The inspectors reviewed the updated LER submittal. No additional performance deficiency or violation of NRC requirements was identified as a result of this review. This LER is Closed.
- (9) LER 50-296/2025-002-00, Unit 3, Reactor Core Isolation Cooling Inoperable due to a Failed Ramp Generating Signal Converter (ADAMS Accession No.
ML25293A446). The inspectors determined that this LER was related to a finding previously dispositioned as NCV 05000296/2025003-01 in Integrated Inspection Report 05000259/2025003, 05000260/2025003, and 05000296/20252003 (ADAMS Accession No. ML25349B313). No additional findings or violations were identified as a result of this LER review. This LER is Closed.
INSPECTION RESULTS
Assessment 71152S Semi Annual Trend in Public Address Speaker Reliability In 2025, approximately eight CRs were associated with PA speaker functionality and reliability. Notably, CR 2041138 documented an adverse trend for PA speaker issues, potentially signaling recurring deficiencies in the public address system. CR 2048748 highlighted actions to address PA speaker concerns identified during a site review, reinforcing the importance of reliable communication systems. CR 2036733 and CR 2035762 further illustrate that PA speaker reliability issues were observed across different contexts, including operational readiness and safety oversight activities.
These reports collectively indicate that PA speaker-related concerns might not be isolated events but part of a broader pattern warranting attention to ensure dependable communication systems. For all identified CRs, corrective actions were initiated, demonstrating recognition of the importance of maintaining PA system integrity for plant safety and emergency preparedness. Further, a root cause evaluation for CR 2032295, written for PA speaker issues at the Watts Bar Nuclear Plant, has an action for the licensee to review the speaker system at Browns Ferry Nuclear as part of the extent of condition review for a White finding.
Failure to Perform Preventive Maintenance Leads to Dual Reactor Recirculation Pump Trips Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000296/2025004-01 Open/Closed None (NPP)71153 A self-revealed Green finding was identified when the licensee failed to classify the auxiliary power distribution contactor for the Unit 2 reactor recirculation pump variable frequency drives as a critical component in accordance with licensee procedure NEDP-12, Equipment Failure Trending, Revision 3. Specifically, the licensee misclassified the contactor as non-critical and this misclassification resulted in no PMs being performed on the component until its age-related failure caused a reactor trip. Had the licensee classified the contactor as critical, PMs would have been procedurally developed and performed to identify and correct degradation of the contactor prior to its failure.
Description:
On May 30, 2025, cooling for both Unit 2 reactor recirculation pump variable frequency drives (VFDs) was lost causing a trip of both pumps and a subsequent manual reactor scram. No safety-related mitigation equipment was lost, and the plant was transitioned to a stable shutdown condition. Troubleshooting found a non-safety-related contactor in the 120VAC auxiliary power distribution lost automatic transfer capability due to an age-related failure. Further review identified that this contactor had no PM history since it was installed in 2004 despite licensee programmatic requirements for performing PMs on this component.
In 2004, the licensee replaced the motor generator drive sets with solid state variable frequency drives under design change notice 50869B, Unit 2 Reactor Recirculation Pump Variable Frequency Drive (VFD), dated March 13, 2003. This design change notice identified 2-CONT-068-002, Contactor for VFD 2A and 2B 120VAC Aux Power Distribution, as a single point failure vulnerability in accordance with NEDP-12, Equipment Failure Trending, Revision 3, because a contactor failure would cause a reactor trip.
NEDP-12 states, in part, that a critical component is a component failure that would result in a Single Point Failureper DS-E2.0.2. Design specification DS-E2.0.2, Single Point Failure for Power Generation Reliability, dated August 22, 2001, states, in part, that a single point failure is a single component or portion of a system whose failure will result in areactor trip. SPP-6.2, Preventive Maintenance, Revision 3, defines a preventive maintenance critical component (PMCC) as one whose failure would result in a plant trip. SPP-6.2 requires an evaluation of PMCCs to determine the appropriate PMs to assure that degradation will be identified and corrected before malfunction occurs.
In or around May 2004, the licensee incorrectly categorized this contactor as non-critical due to a human performance error. This error resulted in this component not being evaluated for appropriate PMs as per SPP-6.2. Consequently, no PMs were either developed or performed on the contactor. In the root cause evaluation for this issue, the licensee concluded that if the contactor had been appropriately classified, mitigation actions would have included a two-year functional test and a ten-year replacement PM.
The inspectors concluded that the failure to properly classify the component as PMCC directly led to the failure to develop any PMs to identify and correct degradation of the contactor prior to its failure in 2025. Had the licensee performed PMs on the contactor, either the degradation likely would have been identified and corrected before it failed or the contactor would have been replaced on a nominal ten year frequency - well before the in-service failure time of 21 years.
Corrective Actions: The licensee replaced the failed contactor, classified the contactors as critical, and developed PMs for all three units.
Corrective Action References: CR 2017005
Performance Assessment:
Performance Deficiency: The licensees failure to classify the auxiliary power distribution contactor for the Unit 2 reactor recirculation pump variable frequency drives as a critical component in accordance with licensee procedure NEDP-12 was a performance deficiency.
Specifically, the licensee misclassified the contactor as non-critical and this misclassification resulted in no PMs being performed on the component until its age-related failure caused a reactor trip.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to perform PMs led to an age-related contactor malfunction. This contactor malfunction caused the loss of both reactor recirculation pumps and a reactor trip.
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 1 of Appendix A, the inspectors answered No to question B, Transient Initiators. Although the finding caused a scram, it did not cause the loss of any mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition; therefore, the finding screened to Green.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
Pressure Boundary Leaks on Reactor Recirculation Small Bore Piping Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000260/2025004-02 Open/Closed None (NPP)71153 A self-revealed Green NCV of TS 3.4.4, "Reactor Coolant System Operational Leakage," was identified when the licensee failed to accurately model the cyclic forces for the small bore piping associated with the Unit 3 reactor recirculation system. As a result, a pressure boundary leak developed on two small bore reactor recirculation lines when flow-induced vibrations caused cyclic fatigue failure of both lines.
Description:
On October 5, 2024, during a drywell walkdown for Unit 3 planned outage F313, Browns Ferry Nuclear Plant (BFN) personnel discovered two through-wall piping leaks on two pipe assemblies associated with both reactor recirculation pumps. Both leaks were due to socket weld cracks in a 1-inch pressure sensing line for the 3B reactor recirculation pump and a 3/4-inch vent line elbow for the 3A reactor recirculation pump. Both assemblies were classified as ASME Code Class 2 piping and constituted part of the Unit 3 reactor coolant system pressure boundary. The licensee submitted LER 05000296/2024-004-00 (ADAMS Accession No. ML24339B794) on December 4, 2024.
This condition is assumed to have developed on June 3, 2024, when system monitoring detected an increase in unidentified drywell leakage and ended on October 5, 2024, when Unit 3 was shut down. The licensee replaced the affected components and implemented a design change on similar piping for all BFN Units. The licensee determined that the cracks were caused by high-cycle fatigue that exceeded the endurance limit.
The direct cause of the failures in both lines was cracking in the socket welds, with high-cycle fatigue identified as the apparent cause. The fatigue cracking was primarily driven by operating stresses that were intensified by plant power uprates and the implementation of Maximum Extended Load Line Limit Analysis Plus, which altered reactor recirculation pump speeds. These operating changes increased vibrational loads and increased the likelihood that piping systems would be excited at resonant frequencies. During evaluations performed for the extended power uprate in 2007, the small bore piping was assumed to be rigid and was therefore not modeled independently from the large bore piping. As a result of this modeling approach, the licensee did not identify that flow-induced vibration stresses under normal operating conditions were sufficient to exceed the endurance limit of the affected small bore pipe assemblies.
Corrective Actions: The licensee replaced the failed 3/4-inch and 1-inch small bore piping elbows to restore code class boundary integrity. In addition, design modifications are being implemented on the 3A and 3B recirculation pump vent, seal vent, and associated sensing line piping for all three Browns Ferry units to reduce susceptibility to flow-induced vibration and high-cycle fatigue. These actions are intended to mitigate vibrational stresses, prevent recurrence of socket weld cracking, and ensure long-term structural integrity of the affected ASME Code Class 2 piping.
Corrective Action References: CR 1963793 and 1963794
Performance Assessment:
Performance Deficiency: The failure to accurately model the small bore piping in the reactor recirculation system in accordance with 10 CFR 50, Appendix B, Criterion III, Design Control, was a performance deficiency. Specifically, the licensee failed to accurately calculate the stresses on the failed pipe assemblies. This failure ultimately led to a pressure boundary leak which could have been predicted and prevented with accurate modeling.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the reactor coolant system (RCS) Equipment and Barrier Performance 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 licensee failed to accurately calculate the stresses on the failed pipe assemblies. This failure ultimately led to two pressure boundary leaks which could have been predicted and prevented with accurate modeling.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding screened out in the review of the Barrier Integrity cornerstone as the PD was not related to pressurized thermal shock; therefore, the finding was addressed under the Initiating Events cornerstone. Since a reasonable assessment of degradation, could have resulted in exceeding the RCS leak rate for a small break loss of coolant accident (LOCA), a detailed risk evaluation was performed by a regional Senior Risk Analyst (SRA) in accordance with IMC 0609, Appendix A, utilizing the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) 8, Version 8.2.12, and the NRC Browns Ferry Unit 3 Standardized Plant Analysis Risk (SPAR) model, Version 8.82, dated August 12, 2023. The exposure period was from June 3, 2024, when indications of the leak were present until October 5, 2024, when the plant was taken to Mode 4, a total of 124 days. The PD was conservatively modeled as an increase in the small break LOCA frequency by two orders of magnitude given the leak of the vent line had the potential to cause the socket to fail and initiate an unisolable small break LOCA if the entire socket failed and a one order of magnitude increase in medium break LOCA frequency to account for the possibility that both sockets could fail creating an equivalent 1-1/4 inch break size worst case. This is conservative since plant technical specifications require a plant shutdown when unidentified leak rates is greater than 5 gallon per minute or there is a change in unidentified leak rate greater than 2 gallons per minute in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. The dominant sequence was a medium break LOCA with a failure of high pressure injection and operators failing to manually depressurize the reactor. This bounding detailed risk evaluation estimated that the PD resulted in an increase in core damage frequency of <1.0 E-6/year, a GREEN finding of very low safety significance.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: TS LCO 3.4.4, "Reactor Coolant System Operational Leakage," requires, in part, that operational leakage shall be limited to no pressure boundary leakage while in Modes 1, 2, and 3; otherwise, the unit shall 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 Mode 4 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 June 3, 2024, to October 5, 2024, Unit 3 operational leakage included pressure boundary leakage while in Mode 1, the unit was not 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 Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Main Steam Relief Valves Lift Settings Outside of Technical Specification Required Setpoints Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000260/2025004-03 Open/Closed Not Applicable 71153 A self-revealed Severity Level IV NCV of TS 3.4.3, "Safety/Relief Valves (S/RVs)," was identified when the licensee discovered, through as-found test results, that 4 of 13 main steam relief valves (MSRVs) had as-found lift settings outside of the +/-3 percent setpoint band required for their operability.
Description:
BFN Unit 2 TS 3.4.3, "Overpressure Protection System," requires 12 of 13 MSRVs to be operable while in Modes 1, 2, and 3. On June 25, 2025, the licensee was notified of as-found testing results that four Unit 2 MSRVs were outside of the +/-3 percent setpoint band required for operability per TS surveillance requirement 3.4.3.1. It was determined that all four MSRVs failed due to corrosion bonding to the valve seat. All four MSRVs were considered to be inoperable during the entire operating cycle from March 21, 2023, to March 5, 2025, and longer than permitted by TS 3.4.3.
The affected valves remained capable of maintaining reactor pressure below the ASME code limit of 1375 psig. All 13 of the MSRV pilot valves were replaced with pre-certified spares during the Unit 2 Spring 2025 refueling outage. The previous corrective action from LER 05000260/2021-002-00 to apply a platinum coating to the pilot valve using the plasma enhanced magnetron sputtering (PEMS) deposition method, which improves the quality and adhesion of the coating, has been utilized. A flaking issue has been noted with the platinum coated pilot disc. The Boiling Water Reactor Owners' Group is continuing to work toward a solution to improve the quality and adhesion of the platinum coating on the discs.
Corrective Actions: The licensee replaced all 13 MSRV pilot valves during the Spring 2025 refueling outage. As-left testing verified that these refurbished pilot valves were within +/- 1 percent of their setpoints. The installed valves have implemented corrective actions from past occurrences of corrosion bonding that include preparing the pilot discs in accordance with the revised procedure and vendor recommendations. The currently installed refurbished valves had platinum coatings applied utilizing the PEMS deposition method, and as-left values were verified to be within +/- 1 percent of their setpoints. Additional corrective actions may be developed based on feedback from the Boiling Water Reactor Owners Group related to corrosion bonding of this specific model safety relief valve.
Corrective Action References: CRs 1985102 and 2022214
Performance Assessment:
The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.
Enforcement:
Severity: This violation is characterized as a Severity Level IV violation based on its similarity to SL IV example 6.1.d.1 in the Enforcement Policy. The inspectors also reviewed NRC Enforcement Policy, Section 2.2.1, "Factors Affecting Assessment of Violations," which states, in part, that in determining the appropriate enforcement response to a violation, the NRC considers, whenever possible, risk information in assessing the safety or security significance of violations and assigning severity levels. The inspectors determined the issue to be of very low safety significance because the valves remained capable of performing their required safety function.
Violation: BFN Unit 2 TS 3.4.3, "Safety/Relief Valves," Condition A, requires that with one or more required S/RVs inoperable, the unit 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 Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Contrary to the above, four required S/RVs were inoperable from March 21, 2023, to March 5, 2025, and the unit did not enter Mode 3 and Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Licensee-Identified Non-Cited Violation 71153 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: BFN Unit 2 TS LCO 3.6.2.4, "Residual Heat Removal (RHR) Suppression Pool Spray (SPS)," requires while in Modes 1, 2, and 3 that four RHR SPS subsystems shall be operable. TS LCO 3.6.2.4 Condition B requires, with two RHR SPS subsystems inoperable, that one RHR SPS subsystem be restored to OPERABLE status within seven days. If the required actions for Condition B are not met within the established completion time, Condition D requires the unit to be in Mode 3 and Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively.
Contrary to the above, on multiple occasions from March 20, 2019, to March 24, 2025, two RHR SPS subsystems were inoperable while Unit 2 was in Modes 1, 2, and 3. e subsystems were inoperable for 7 days while in Mode 1, 2, and 3 BFN Unit 2 TS LCO 3.8.1 Condition B requires while in Modes 1, 2, and 3, with one required Unit 1 and 2 standby diesel generator (SDG) inoperable, that required feature(s) supported by the inoperable Unit 1 and 2 SDG be declared inoperable when the redundant required feature(s) are inoperable at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from the discovery of a condition concurrent with inoperability of the redundant required feature.
Contrary to the above, on multiple occasions from March 20, 2019, to March 24, 2025, the RHR "B" or "D" suppression pool spray loops were not declared inoperable when the "C" or "D" SDGs were inoperable for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while Unit 2 was in Modes 1, 2, and 3.
Browns Ferry Nuclear Unit 2 TS LCO 3.0.4 states 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 six occasions between April 9, 2019, and March 22, 2023, Unit 2 entered a TS 3.6.2.4 Applicable Mode when TS 3.6.2.4 Required Actions were not met.
Specifically, Unit 2 RHR suppression pool spray Loop I was inoperable while entering Modes 2 and 1 (e.g. starting up following a unit shutdown) as stated above.
Significance/Severity: Green. The inspectors used IMC 0609, Appendix A, Exhibit 2 "Mitigating Systems Screening Questions" to assess the significance. Because neither the licensee nor the NRC PRA models credit suppression pool spray (only drywell spray) for over pressure control, the SSC maintained its PRA functionality. The inspectors answered "YES" to question A.1 and the issue screens to GREEN.
Corrective Action References: CR 2000493
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On January 8, 2026, the inspectors presented the integrated inspection results to Daniel Komm, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
LOR-EXAM-83
Simulator Exercise Scenario
Work Orders
23458722
Corrective Action
Documents
1979541, 1461550, 1916891, 1805633,
2017005, 1681171, 1668815, 1110412,
2017921, 1962724, 1954970, 1954800,
1954797, 1917265, 1912342, 1912331
Engineering
Evaluations
Maintenance Rule Get Well Plan
EHPM Function 7-D
Miscellaneous
System 068 Reactor Recirculation
Pump VFDs (a)(1) Determination
Evaluation
8/25/2025
Miscellaneous
U1/2/3 System 068, Reactor
Recirculation VFDs, (a)(1) Plan
Miscellaneous
U1/2/3 CR 105 Contactors (a)(1) plan
Miscellaneous
U1/2/3 CR 105 Contactors (a)(1) plan
Miscellaneous
DS-E2.0.2
Design Specification for Single Point
Failure for Power Generation Reliability
Work Orders
25017724, 122455138
Corrective Action
Documents
2040968, 2041571, 2044150, 2051196,
2052912, 2054399
Drawings
1-47B458-898
Mechanical Core Spray Cooling
System Pipe Support
Drawings
1-47B458-899
Mechanical Core Spray Cooling
System Pipe Support
Drawings
2-47E811-1
Flow Diagram Residual Heat Removal
System
Work Orders
24678560, 123797028, 125788612,
25790006, 125790731
Corrective Action
Documents
1598981, 2042199
Work Orders
24528340, 124496142, 121257428,
24772979, 125328455
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
2041138, 2036733, 2032295, 2035762,
2048748
Corrective Action
Documents
CR 2017005, 2033486, 2033773, CR
2034237, 2000493
Work Orders
25556690, 125559901, 125556569