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| Entered date | Event description | |
|---|---|---|
| ENS 56257 | 3 December 2022 13:05:00 | The following information was provided by the licensee via email: On 12/2/2022 at 2330 (CST) during the planned F311 outage on Browns Ferry Nuclear Plant Unit 3, personnel entered the Unit 3 drywell for leak identification. Personnel discovered a through-wall piping leak on a 0.75 inch test line between the two test line isolation valves. This 0.75 inch test line is located on the residual heat removal (RHR) loop 1 shutdown cooling and RHR return line to the reactor vessel. On 12/3/2022 at 1000 CST, Engineering determined this location is classified as ASME Code Class 1 piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. |
| ENS 53462 | 20 June 2018 11:39:00 | On June 20, 2018 at 1003 CDT, the licensee declared a Notification of Unusual Event based on Emergency Action Level (EAL) 6.5.U, toxic gas release on site. The Notification of Unusual Event was terminated at 1025 CDT. The toxic gas release occurred when site personnel were filling a fire suppression carbon dioxide (CO2) tank outside the diesel generator building. The relief valve in the common diesel generator room for Unit 1 and 2 diesel generators inadvertently lifted causing a toxic gas environment by releasing CO2 into the room. The licensee terminated the tank fill stopping the release of CO2, and with the door to the room being opened, the gas cleared in about 20 minutes. The licensee has notified the NRC Resident Inspector. Notified DHS SWO, FEMA Ops, DHS NICC, FEMA NWC (email) and NuclearSSA (email). |
| ENS 53300 | 29 March 2018 22:28:00 | At 1344 on March 29, 2018, it was determined (engineering evaluation) that an unanalyzed condition that significantly degraded plant safety previously existed. During a postulated control room abandonment due to a fire, and concurrent with a Loss of Offsite Power (LOOP), the required number of Emergency Equipment Cooling Water (EECW) pumps would not have been available from 10/28/2015 to 3/10/2018. On March 8, 2018, during relay functional testing it was discovered that the C3 Emergency Equipment Cooling Water (EECW) pump closing springs did not recharge with the breaker transfer switch in emergency. On August 23, 2012, a wire modification was performed that contained a drawing error resulting in wire placement on the incorrect connection points for the C3 EECW pump. On March 10, 2018, the C3 EECW pump breaker wiring was corrected and subsequent testing was completed satisfactorily. Prior to 10/28/2015, Brown's Ferry Nuclear Plant (BFN) adhered to Appendix R fire protection requirements which did not credit the C3 EECW pump for fire protection from the backup control location. On 10/28/2015, BFN transitioned to National Fire Protection Association (NFPA) 805 fire protection requirements which takes credit for the C3 EECW pump from the backup control location. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B), 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety'. This is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(ii)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified of this event. |
| ENS 52648 | 29 March 2017 23:36:00 | At 1844 CDT on 3/29/2017, Unit 2 initiated a manual scram due to multiple rods inserting. At 1842 during Unit 2 start-up, Intermediate Range Monitor (IRM) 'G' drifted low. The operator adjusted the range down one position with no immediate reaction. At 1844, a spike on IRM 'G' caused a half scram on Reactor Protection System (RPS) 'A' trip system. The half scram was being reset after evaluating no trip condition was present. As the operator reset groups 2 and 3, a trip signal from IRM 'F' was received on the RPS 'B' trip system, resulting in rod insertion for groups 1 and 4. When the operator identified multiple rods inserting, the actions of procedure 2-AOI-100-1 were followed and a manual scram was inserted. Investigation is ongoing. All safety systems remained in standby readiness configuration. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary Containment Isolations Systems did not receive an actuation signal and performed as designed. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the RPS when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of systems listed in paragraph (b)(3)(iv)(B) Reactor Protection System(RPS) including reactor scram and reactor trip'. This event requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. |
| ENS 52557 | 16 February 2017 15:57:00 | On February 16, 2017 at 1052 CST, Unit 2 received a High Pressure Coolant Injection (HPCI) System 120V Power Failure alarm. Troubleshooting identified a cleared fuse for the HPCI System Flow Controller, 2-FIC-73-33, which would have prevented automatic or manual HPCI System initiation and rendered the HPCI System inoperable. At 1145 CST, the cleared fuse was replaced and the HPCI system was declared available. The HPCI System remains inoperable for additional troubleshooting. This constitutes an unplanned HPCI System inoperability and requires an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(v)(D), due to the failure of a single train system affecting accident mitigation and a 60 day written report in accordance with 10 CFR 50.73(a)(2)(v)(D). The Senior NRC Resident Inspector has been notified. |
| ENS 51878 | 22 April 2016 18:22:00 | At 1359 CDT on April 22, 2016, Browns Ferry Units 1 & 2 experienced a partial loss of power during the transfer of Shutdown Bus 2 from the alternate power source back to the normal power source. During the transfer, the normal feeder breaker failed to close and provide power to the Shutdown Bus, resulting in an auto actuation of two Emergency Diesel Generators (EDGs). Power to Shutdown Bus 2 was immediately restored using the alternate feeder breaker. The EDGs did not tie to the boards. All systems responded as expected for the loss of power, and both Units 1 & 2 maintained 100% Rx Power. All systems have been restored to a normal lineup, and both Units 1 & 2 remain at 100% Rx Power. This event requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified. The cause of the normal feeder breaker failure is being investigated. There was no impact on Unit 3. |
| ENS 51571 | 29 November 2015 20:40:00 | On November 29, 2015 at approximately 1600 Central Standard Time, it was reported that the oil catch device in place around the Temporary Diesel Generators (TDG) had overflowed due to a combination of accumulated rainwater and failure of one wall of the catch device to float. There is a small oil leak from the TDG. Less than one gallon of oil is estimated to have been spilled. The oil spill has reached the BFN (Brown Ferry Nuclear) forebay (waters of the US) via the station's storm drainage as evidenced by an oil sheen on the water of the forebay. The source of leakage has been contained. This oil spill is reportable to the EPA (National Response Center) under 40CFR112. Notification to the National Response Center has been made. Additional notification has been made to Alabama Emergency Management. The licensee notified the NRC Resident Inspector. |
| ENS 51153 | 12 June 2015 17:18:00 | On June 12, 2015 at 1030 CDT, the Browns Ferry Nuclear Plant Unit 3 High Pressure Coolant Injection (HPCI) system was declared inoperable due to the time to drain the Turbine Exhaust Drain Pot after running the system for periodic testing. The concern is that the turbine may be partially flooded after shutting down and a subsequent restart could cause a water hammer event, possibly damaging the system. This issue was previously analyzed by Engineering as acceptable, but the time to drain the pot after the latest test indicates more water in the exhaust than the maximum amount used in the analysis. Technical Specification 3.5.1, ECCS Operating, Condition C, was entered as a result of the inoperable HPCI system. This constitutes an unplanned HPCI system inoperability and requires an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(v)(D) due to the failure of a single train system affecting accident mitigation, and a 60 day written report in accordance with 10 CFR 50.73(a)(2)(v) The NRC Resident Inspector has been notified. The Technical Specification Action statement allows 14 days to restore the HPCI system to operable status.
Browns Ferry Nuclear Plant is retracting the 8-hour NRC notification (EN# 51153) made on June 12, 2015 at 1030 CDT. The notification on June 12, 2015, reported a condition where the HPCI system was declared inoperable due to the time to drain the Turbine Exhaust Drain Pot after running the system for periodic testing. The concern was that the turbine may be partially flooded which could cause water hammer and damage the HPCI system. Subsequent evaluation concluded that the HPCI system under the identified flooded turbine conditions will not produce a transient that exceeds design values, therefore, HPCI system operability was maintained and no reportable condition existed during this time. The NRC Resident Inspector has been notified of this retraction. Notified R2DO (Shaeffer). |
| ENS 50725 | 8 January 2015 18:14:00 | On January 7, 2015, at approximately 0700 CST, a leak to the environment was identified. Tritium was present at a concentration of 7.52E-3 uCi/mL, which is above the US Environmental Protection Agency (EPA) drinking water standard of 20,000 picocuries per liter. No other radioactive isotopes were identified. The leak rate was estimated at approximately 0.5 gpm and determined to be from the condensate head tank. Water was accumulating on the concrete-lined reactor/refuel air zone air intake plenum with some accumulation of water on the ground in the area. The intake plenum contains three floor drains. Actions were immediately taken to terminate the leak once the flowpath was identified. The flowpath was terminated two hours and 45 minutes after identification. Based on system review and analysis, any tritiated water that would have made it to the floor drains would then be mixed with incoming raw water at two million gallons per minute. This mixed volume of water would then be circulated through the plant and discharged to the river with a resultant tritium concentration that is much less than detectable levels and well below US EPA drinking water standards. The station has established increased monitoring of groundwater at designated sample wells. On January 8, 2015, at approximately 1645 CST in accordance with TVA procedures and the guidance of NEI 07-07 (Nuclear Energy Institute) for the Groundwater Protection Initiative, the licensee notified the Alabama Radiological Protection Department and Alabama Department of Environmental Management. The Limestone County Emergency Management Department will be notified. The licensee notified the NRC Resident Inspector. |
| ENS 50719 | 6 January 2015 17:27:00 | On January 6, 2015 at approximately 1227 CST, a visible oil sheen was reported from a TVA owned boat that sank next to the dock on the Tennessee River near Cooling Tower 1. This oil spill is reportable to the EPA (National Response Center) under 40CFR112. Notification to the National Response Center will be made. This oil sheen was determined to be more than minor to enter US waters. The source of the oil sheen has been contained using oil booms. Additional notifications will be made to Alabama Emergency Management. The licensee has notified the NRC Resident Inspector. The licensee notified the Environmental Protection Agency and the National Response Center. |
| ENS 50090 | 6 May 2014 13:27:00 | At 0830 (CDT) on 05/06/2014, the Unit 3 reactor automatically scrammed due to low reactor water level as a result of a trip of both recirculation pumps. Main Steam Isolation Valves remained open with main turbine bypass valves controlling reactor pressure. Reactor feedwater pumps are in service to control reactor water level. Primary Containment Isolation System Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. The Reactor Feedwater System controlled and maintained water level above the level 2 initiation setpoint. Prior to the Scram, the reactor was operating at 100% power. A Core and Containment Cooling Systems Analog Trip Unit Functional Test was in progress. The cause of the recirculation pump trip is under investigation. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. U1 and U2 remained at 100% power and were unaffected.
Investigation revealed that a failed power supply caused an Anticipated Transient Without Scram/Alternate Rod Insertion (ATWS/ARI) signal to be generated when a level 2 Reactor Water Level was simulated on one instrument. All systems responded to the ATWS/ARI signal as designed. This signal opened the Recirc Pump Trip breakers for both Recirculation Pumps and opened the ARI valves to bleed air from the Reactor Protection System (RPS) scram air header. The resulting transient caused reactor water level to dip below the RPS trip setpoint (level 3 Reactor Water Level), a normal plant response, and the automatic scram signal occurred. At the time of the RPS scram signal, all rods were inserting and reactor power was approximately 2-3% and lowering. The NRC Resident Inspector has been notified. Notified R2DO (Bonser). |
| ENS 49455 | 19 October 2013 23:00:00 | On 10/19/2013 at 1705 CDT, Browns Ferry Nuclear Plant (BFN) experienced a partial loss of offsite communications, which included the NRC Emergency Notification System, the Emergency Plan Paging System and normal phone communication to offsite. Main control room indications and assessment capability were not affected. The Nextel cellular phone system and the satellite phones in the control rooms were unaffected. The plant radio system was also unaffected. The Emergency Response Data System is currently out of service due to planned maintenance on the associated offsite power supply. Operation of BFN Unit 1, Unit 2, and Unit 3 was not affected by the event. The BFN facilities personnel found a blown fuse in one of the temporary Diesel Generators (DGs) supplying the node associated with the outside phone communications. The fuse was replaced and the DG has been in-service with no issues. Normal communications and access have been restored. The NRC Senior Resident Inspector has been notified. This event is reportable under 10CFR50.72(b)(3)(xiii), 'Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability' (e.g., significant portion of control room indication, emergency notification system, or offsite notification system). |
| ENS 49382 | 24 September 2013 23:21:00 | On 9/24/13 at 1530 CDT, the Browns Ferry Nuclear Plant Unit 2 High Pressure Coolant Injection (HPCI) System was declared inoperable due to inoperability of the minimum flow valve (2-MVOP-073-0030). Engineering identified during review of EQ (Environment Qualification) WO (Work Order) # 113598388 that the motor leads for 2-MVOP-073-0030 were taped instead of terminated with a Raychem splice. The valve actuator is an EQ component and requires termination of the motor leads to the incoming power cable by Raychem splice or a Marathon 300 terminal block. Engineering evaluation was requested and operability of 2-MVOP-073-0030 could not be supported due to the tape being unanalyzed for harsh environmental conditions. The HPCI Minimum Flow valve has a required OPEN safety function to prevent overheating the HPCI pump and a CLOSED safety function to provide containment isolation. Technical Specification 3.5.1, Emergency Core Cooling System-Operating, Condition C was entered for HPCI system inoperability. In addition, the actions of Technical Specification 3.6.1.3, Primary Containment Isolation Valves, were entered due to the inoperability of the primary containment isolation function of the HPCI Minimum Flow valve. This incident is reportable as an 8-hour ENS notification under 10 CFR 50.72 (b)(3)(v)(D) as any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. It also requires a 60 day written report in accordance with 10 CFR 50.73(a)(2)(vii). The NRC Resident Inspector has been notified. SR number associated with this report: 784462 The licensee is in a 14-day LCO action statement to return HPCI to operable under TS 3.5.1. The LCO for TS 3.6.1.3 was satisfied. |
| ENS 47374 | 25 October 2011 23:49:00 | At 1534 on 10/25/11 Site Engineering and Licensing reported the following to the Operations Department: The following deficiency was identified during reviews for NFPA 805 Transition: Cable 1B11 (#12 awg - Fire Area Route 16, 17) from the shunt in the 250 VDC Battery Board 1 (0-BDDD-280-0001) compartment #1, to the control room ammeter 1-EI-57-38 mounted on Control Room Panel 1-9-8, is not fused or otherwise protected against an electrical fault. The 250 VDC Battery 1 is an un-grounded DC system. However, in a scenario where the 250 VDC Battery Board system negative becomes grounded at the same time the ammeter cable 1B11 has a fault to ground, the cable may not be protected and the cable could auto-ignite anywhere along the cables length from the ground fault location back to the battery source. For example, in a Fire Area 16 (Control Bay) scenario, these ground faults could result in a fire being spread from Fire Area 16 to Fire Area 17 (Battery Board 1). Fuses should be added to protect these cable conductors. A similar condition exists for the remote ammeters on Unit Battery Board 2 (Cable 2B11 - Fire Area Route 16, 18), Unit Battery Board 3 (Cable 3B11 - Fire Area Route 16, 19), Unit Battery Board 4 (Cable 2B4018 - Fire Area Route 16, 26), Unit Battery Board 5 (Cable 1B383 - Fire Area Route 16, 26), and Unit Battery Board 6 (Cable 3B336 - Fire Area Route 16, 26). This condition is reportable as an 8 hour notification to the NRC IAW 10CFR 50.72(b)(3)(ii)(B), 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' This is also reportable as 60 day written report IAW 10CFR 50.73(a)(2)(ii)(B). Fire watch compensatory measures are in place for the Fire Areas of concern. Ref. FPIP#09-1920 The NRC resident has been notified of this event. This event was entered into the licensee's Corrective Action Program as PER# 452185. |