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| Entered date | Event description | |
|---|---|---|
| ENS 55287 | 1 June 2021 17:46:00 | This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of the 2A Reactor Protection System (RPS). On April 1, 2021, at 1302 (CDT), Browns Ferry Unit 2, 2A RPS (Motor Generator) MG set tripped causing a half scram. Unit 2 experienced an unexpected trip of the 2A RPS MG Set that resulted in automatic Primary Containment Isolation System (PCIS) Group 2, 3, 6, and 8 isolations and Trains A, B, and C Standby Gas Treatment (SGT) and Train A Control Room Emergency Ventilation (CREV) starts. At the time of the event, Unit 2 was in a refueling outage and the rods were already fully inserted. All systems responded as expected. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. Based on the troubleshooting conducted, the cause was determined to be a loose wiring connection in the motor circuit. The lugs were replaced with ring lugs. Operations reset the 2A RPS Half Scram and PCIS in accordance with 2-AOI-99-1 on April 1, 2021, at 1324 CDT thus correcting the condition and returning RPS to service. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1683358. The NRC Resident Inspector has been notified of this event. |
| ENS 54795 | 21 July 2020 08:58:00 | The following was received from TVA - Brown's Ferry at 0858, 21 July 20. On July 21, 2020, at 0435 hours Central Daylight Time, Browns Ferry Unit 1 inserted a manual reactor scram due to degrading main condenser vacuum from marine biofouling at the intake structure. Browns Ferry Unit 2 is in Mode 4 and Browns Ferry Unit 3 is at approximately 76% rated thermal power and stable. Primary Containment Isolation Systems received an actuation signal for groups 2, 3, 6, 8 on reactor water level below +2". All Primary Containment Isolation System groups that received an actuation signal performed as designed. Additionally, all other systems functioned as designed. This event is reportable within 4 hours per 10CFR50.72(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. This event also requires an 8 hour report per 10CFR50.72(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), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation and (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). The NRC Resident Inspector has been notified. The plant is stable in Mode 3 and will remain shutdown until marine growth clogging the intake structure abates. |
| ENS 54794 | 20 July 2020 18:11:00 | On July 20, 2020, at 1325 hours Central Daylight Time, Brown's Ferry Unit 2 inserted a manual reactor scram due to degrading main condenser vacuum from marine biofouling at the intake structure. Brown's Ferry Unit 1 performed a down power to 43% and Unit 3 down powered to 76%. Conditions are stable on both Unit 1 and 3 following unit down power. Primary Containment Isolations Systems received an actuation signal for groups 2, 3, 6, and 8 on reactor water level below +2". All Primary Containment Isolations System groups that received an actuation signal performed as designed. Additionally, all other systems functioned as designed. This event is reportable within 4 hours per 10CFR50.72(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. This event also requires an 8 hour report per 10CFR 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), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation and (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). The NRC Resident Inspector has been notified. All control rods fully inserted and decay heat is being removed via normal feedwater and condenser. |
| ENS 54162 | 12 July 2019 22:50:00 | At 1640 CDT on 7/12/19, Unit 1 High Pressure Coolant Injection (HPCI) received an invalid auto isolation signal which closed the HPCI steam supply valves rendering HPCI inoperable. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D), as an 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. The isolation occurred while performing a calibration and functional check of a level switch for the Unit 1 Core Spray system. Continuity was checked across the incorrect set of contacts which completed the circuit in logic bus 'A' for the auto isolation signal in the HPCI system. There was no impact to the safety of the public or plant personnel during the time HPCI system was isolated. HPCI was returned to operable at 2110 CDT on 7/12/19. CR 1532094 documents this condition in the Corrective Action Program. The licensee has notified the NRC Resident Inspector |
| ENS 53678 | 21 October 2018 06:19:00 | At 0200 Central Daylight Time on 10/21/2018, Browns Ferry Nuclear Plant Unit 3 commenced a reactor shutdown as required by the Technical Requirements Manual Limiting Condition for Operation 3.4.1 Coolant Chemistry Condition D due to conductivity greater than 10 micro mho/cm at 25 degrees Celsius. The required action for this condition is to immediately initiate an orderly shutdown and be in Mode 4 as rapidly as cooldown rate permits. This event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified.
ENS Event Number 53678, made on 10/21/18, is being retracted. NRC notification 53678 was made to ensure that the four-hour non-emergency reporting requirements of 10 CFR 50.72 were met when the licensee discovered a condition requiring shut down of a reactor. 10 CFR 50.72 requires a report in accordance with 50.72(b)(2)(i) for any Technical Specifications (TS) required reactor shutdown. NUREG-1022 only specifies TS applicability and makes no mention of a Technical Requirements Manual (TRM) required shutdown. Because the shutdown comes from the TRM and not the TS as discussed in 10 CFR 50.72 and NUREG-1022, an EN was not required. TVA's evaluation of this event notification is documented in the corrective action program. The licensee notified the NRC Resident Inspector. Notified the R2DO (Ehrhardt). |
| ENS 53049 | 1 November 2017 22:26:00 | At 1425 (CDT) on November 1, 2017, Operations was notified of a condition affecting Unit 3 4kV Shutdown Boards 3EA, 3EB, 3EC, and 3ED. It was discovered that multiple potential transformer (PT) primary fuses are GE type EJ1 size 0.5 AMP which does not coordinate with the PT's secondary fuses. A fault on the associated cable could clear the primary PT primary fuses for the 4kV Shutdown Board. This would result in the board tripping 4kV motor loads, disconnecting from Off-site power and connecting to the Emergency Diesel Generator. However, since the PT fuse is cleared, the under-voltage trips on the 4kV motors would remain in if there is no Common Accident Signal (CAS) present. The 4kV motor loads include Residual Heat Removal (RHR) Pumps, Core Spray (CS) Pumps, Residual Heat Removal Service Water (RHRSW) Pumps, and Emergency Equipment Cooling Water (EECW) pumps. Review of NFPA 805 analyses show the cables for all four U3 4kV Shutdown Boards are routed in Fire Area 03-03 and Fire Area 16. Therefore a fire in either area could result in a loss of all four U3 4kV Shutdown Boards motor loads. Cables for 4kV Shutdown Board 3EA and 3EB are both routed in Fire Area 21 which could result in a loss of both Division I Shutdown Board motor loads. Compensatory fire watch measures have been established. This event requires an 8 hour report in accordance with 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.' The NRC Resident Inspector has been notified. CR 1354129 was initiated in the Corrective Action Program. |
| ENS 52273 | 30 September 2016 16:19:00 | A licensed operator had a confirmed positive for alcohol on a random fitness for duty test. The employee's access to the plant has been terminated. The licensee has informed the NRC Resident Inspector. |
| ENS 52150 | 3 August 2016 21:38:00 | On August 3, 2016 at approximately 1300 CDT during review of NFPA 805 requirements, it was determined that the NFPA 805 analysis and Fire Safe Shutdown (FSS) procedures do not consider the potential for fire-induced failure of the 4kV Shutdown Board under-voltage trip functions for Emergency Diesel Generator (EDG) power supply alignments. As such, a condition could possibly exist during a postulated fire where a required EDG's 4kV loads would not trip on an undervoltage condition. Current procedures and timeline analysis do not consider operator actions that could be necessary to manually strip the 4kV Safe Shutdown (S/D) board prior to subsequent EDG restart. As such, a subsequent restart, manual or automatic, of the EDG under these conditions, with its associated loads still connected to the 4kV S/D board, could potentially over load the EDG on restart. This notification is to report a condition involving a deficiency in FSS procedures affecting restoration of power to safe shutdown busses under certain postulated fire scenarios. The condition could result in an adverse impact on the ability of operators to implement FSS procedures in response to a postulated fire in 6 fire areas. Therefore, this notification is being made 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. Compensatory fire watches have been established in the affected areas and this deficiency has been added to the current fire protection impairment plan. The (NRC) Resident Inspector has been notified. |
| ENS 52019 | 18 June 2016 05:49:00 | On 6/17/2016, 2-SR-3.3.6.1.6(3) HPCI (High Pressure Coolant Injection) system time delay relay calibration periodic surveillance was being performed. During a section in the procedure a fuse cleared for the logic bus B power at 2135 (CDT). This rendered the HPCI system unable to be manually or automatically initiated. At 2239 the fuse was replaced and the HPCI system was restored to a standby lineup. HPCl is a single train safety system and this notification is being made in accordance with 10CFR50.72 (b)(3)(v)(D). The NRC Resident Inspector has been notified. |
| ENS 51996 | 8 June 2016 17:51:00 | At 0800 CDT on 6/8/2016, 3-SR-3.5.1.7, 'HPCI (High Pressure Coolant Injection) Main and Booster Pump Set Developed Head and Flow Rate Test at Rated Reactor Pressure' periodic surveillance was being performed. As part of the surveillance procedure, HPCI was declared inoperable per Technical Specification (TS) Limiting Condition for Operation (LCO) 3.5.1 Emergency Core Cooling Systems (ECCS) - Operating, Condition C at 0925. At 1037, the HPCI turbine was started and a turbine trip alarm was received as well as large swings observed on the suction pressure indicator and slow turbine response. The control room operator then manually tripped the HPCI turbine due to the abnormal indications received. HPCl is a single train safety system and this notification is being made in accordance with 10 CFR 50.72 (b)(3)(v)(D). The NRC Resident Inspector has been notified. |
| ENS 51398 | 16 September 2015 10:10:00 | At 0200 CDT on 9/16/2015, the High Pressure Coolant Injection (HPCI) System was manually isolated to stop a steam leak from the stem packing of the HPCI Steam Supply Valve 2-FCV-073-0016. The leak occurred following performance of 2-SR-3.6.1.3.5 (HPCI) HPCI System Motor Operated Valve Operability, which cycled 2-FCV-073-0016. No Area Radiation Monitoring (ARM) or PCIS Area High Temperature alarms were received and no automatic isolation setpoints were reached. HPCI was declared inoperable per Technical Specification (TS) Limiting Condition for Operation (LCO) 3.5.1 Emergency Core Cooling Systems (ECCS)- Operating, Condition C. 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 NRC resident inspector has been notified. |
| ENS 50582 | 30 October 2014 14:35:00 | On October 30, 2014, at 1100 EDT, TVA conducted a briefing for government officials and other stakeholders regarding the decision to accelerate the Boone Reservoir annual drawdown after discovery of a sink hole near the base of the embankment and a small amount of water and sediment found seeping from the river below the dam. TVA is continuously monitoring the dam and conducting an investigation to determine the source of the water seepage. The dam is located upstream of all three TVA nuclear sites. There are currently no nuclear plant operability or safety issues, and TVA is assessing the impacts on the plant licensing bases. A press release was issued at approximately 1300 EDT. This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi) due to notification other government agencies and a news release. The NRC Senior Resident Inspector was notified. |
| ENS 50658 | 8 December 2014 12:23:00 | This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of general containment isolation signals affecting containment isolation valves in more than one system. On October 7, 2014, at 2135 (CDT), while in a refueling outage with the reactor non-critical (Mode 5), work activities were in progress that included replacement of an excess flow check valve and execution of a Technical Specification Surveillance Procedure on the Automatic Depressurization System. Subsequent to valving in a level transmitter (LT), water levels in both the variable and reference legs of the LT were disturbed resulting in a Unit 1 full scram and Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolation signals due to receipt of an invalid low reactor water level signal. The PCIS Groups 2, 3, 6, and 8 isolations caused the initiation of Trains A, B, and C of the Standby Gas Treatment System and Control Room Emergency Ventilation Subsystem 'A'. The Reactor and Refuel Zone ventilation fans tripped and the secondary containment dampers isolated. Operations personnel responded to the PCIS initiation, ensured all equipment operated as designed, and placed affected systems back in service. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Problem Evaluation Report 943038. The NRC Resident Inspector has been notified of this event. |
| ENS 49942 | 21 March 2014 14:10:00 | This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On January 21, 2014, at 0746 hours Central Standard Time (CST), during performance of the 3C Emergency Diesel Generator (EDG) post modification test instructions, the EDG was supplying a shutdown board in isochronous mode when the 3B Residual Heat Removal (RHR) pump was started causing the voltage to drop to 2100 volts. At this time, Browns Ferry Nuclear Plant (BFN) Unit 3, received a half scram and Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolation signals as a result of losing the 3B Reactor Protection System (RPS) Motor Generator (MG) set due to a time delay relay failure on under voltage. The PCIS groups 2, 3, 6, and 8 isolations caused the initiation of all three trains of the Standby Gas Treatment (SBGT) system, Control Room Emergency Ventilation (CREV) subsystem 'A', and the Refuel fans tripped and isolated. Operations personnel responded to the PCIS initiation, ensured all equipment operated as designed, and placed affected systems back in service. Plant conditions which initiate PCIS group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The apparent cause for this condition was a failure of a 3B RPS MG set time delay relay due to lack of a preventive maintenance strategy. The vendor manual for the time delay relay did not specify a qualified life. The replacement relay specified a replacement schedule of 10 years. The relay that failed was installed for approximately 13 years. To address this condition, preventive maintenance is being developed for MG set time delay relays. In addition, the only remaining relay, similar to the failed relay, is scheduled be replaced on August 25, 2014, for the 2A RPS MG set. The licensee has notified the NRC Resident Inspector. |
| ENS 49406 | 3 October 2013 21:10:00 | The Browns Ferry Nuclear Plant (BFN) minimum Operations shift staffing was evaluated for response to a fire in the Control Bay that ultimately leads to entry into Appendix R Safe Shutdown Instructions (SSIs). The result of this evaluation revealed that the minimum Operations shift staffing does not provide sufficient staffing to support SSI required staffing levels. In the event of an Appendix R fire, one BFN unit would be without a Senior Reactor Operator to direct the implementation of the time critical manual actions specified in the SSI procedures. On October 3, 2013, this condition was determined to be an 8 hour non-emergency report in accordance with 10 CFR 50.72(b)(3)(ii)(B) since sufficient shift staffing levels to implement the SSI procedures in the event of an Appendix R fire were not provided. Compensatory measures have been implemented to ensure sufficient shift staffing is provided to implement Appendix R SSIs. The event was entered into the licensee corrective action program as Service Request number 788812. The NRC Resident Inspector has been notified. |
| ENS 49208 | 20 July 2013 12:56:00 | During performance of 1-SR-3.3.5.1.3(D), 'High Pressure Coolant Injection System Condensate Header Low Level Switch Calibration and Functional Test,' the High Pressure Coolant Injection (HPCI) system was declared inoperable due to exceeding the allotted 1 hour time frame for instrument inoperability of 1-LS-073-0056A and 1-LS-073-0056B (low condensate storage tank (CST) level HPCI suction transfer switches). The HPCI system was declared inoperable in accordance with technical specification (TS) 3.3.5.1 Condition D.1. Condition D of TS 3.3.5.1 is required to be entered during performance of this surveillance. The surveillance disables the automatic CST to Torus suction path transfer function of the HPCI system by booting relay contacts in order to perform a functional test of referenced switches. The 1 hour completion time of TS 3.3.5.1 required by action D.1 was exceeded due to issues with associated test equipment. This constitutes an unplanned HPCI system inoperability and requires an 8 hour NRC notification per 10CFR50.72(b)(3)(v)(D). HPCI remained available for automatic and manual injection during this time period. The HPCI system was declared inoperable at 0438 CDT on 7/20/13 and operability (CST to Torus suction transfer function) was restored at 0454 CDT on 7/20/13. The NRC resident inspector has been notified. The level switches (suction transfer switches) had been taken out-of-service for calibration and testing at 0338 CDT on 7/20/13. |
| ENS 48904 | 9 April 2013 17:07:00 | On February 11, 2013, at 0613 hours (CDT), the Reactor Core Isolation Cooling (RCIC) system was manually started during a planned Unit 3 reactor shutdown. A Reactor Feedwater recirculation piping separation resulted in the loss of condenser vacuum and subsequent unavailability of the Main Turbine Bypass Valves. The RCIC system was manually started at 9.2" of condenser vacuum in order to control reactor water level in anticipation of loss of Reactor Feedwater Pumps (RFPs) which occurs at 7" of condenser vacuum. Safety Relief Valves (SRVs) were manually operated to maintain reactor pressure. The reactor water level was controlled in the normal band by RCIC, and Reactor Pressure was controlled with a combination of Reactor Core Isolation Cooling (RCIC) system and SRV manual operation. All systems operated as designed and Reactor water level was maintained in the prescribed band. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. RCIC operation was secured at 1449 (CDT) on 2/11/2013. This event is reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A). During a review of operating logs it was identified that this event met reporting requirements and had not been reported. Therefore, this report does not comply with the 8 hour requirement. This condition has been entered into the corrective action program. Additionally, an LER is required within 60 days per 10CFR50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. |
| ENS 47942 | 22 May 2012 07:38:00 | At 0249 CDT on 5/22/2012, Unit 3 reactor automatically scrammed due to de-energization of Reactor Protection System (RPS) from actuation of 3A Unit Station Service Transformer (USST) differential relay 387SA, which resulted in a loss of 500KV power to Unit 3. This relay was picked up during a transfer of 4KV Unit Board 3C from alternate power (161KV) to normal power (3A USST). Investigation is in progress as to the cause of relay actuation. 500KV power was restored through the alternate feeder breakers from 161KV to all Unit 3 4KV Unit Boards successfully. 161KV remained available during the entire event. Loss of 500KV power lasted less than 30 seconds and power has been restored to all safety related boards. All Unit 3 diesel generators successfully started and tied to their respective 4KV Shutdown Boards. All safety systems responded as expected to the loss of 500KV power. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. RCIC was manually started to control reactor water level. Primary Containment Isolation System (PCIS) and PCIS initiation signals for groups 1, 2, 3, 6 & 8 were received as designed. At the time of the scram, High Pressure Coolant Injection (HPCI) system was tagged out for removal of temporary instrumentation following planned maintenance. 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. |