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| Entered date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 57298 | 30 August 2024 18:30:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: At 1051 CDT on 8/30/2024, during transfer of 4KV shutdown bus 1 to support Unit 1 shutdown activities, the alternate feeder breaker failed to close resulting in 4KV shutdown boards 'A' and 'B' experiencing an under voltage condition. This resulted in 'A' and 'B' diesel generators automatically starting and tying to their respective boards. This condition also caused a loss of reactor protection system (RPS) channel 'A' on Units 1 and 2, resulting in invalid actuation of primary containment isolation system Groups 2, 3, 6, and 8. The failure of the board to transfer was identified during preparation for the evolution, contingency actions were prepared and implemented as planned. The breaker failure to close has been corrected and 4KV shutdown bus 1 is energized on alternate. 4KV shutdown boards 'A' and 'B' have been restored to offsite power supplies and the diesel generators are secured. All systems responded as expected for the loss of voltage. This event requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC resident has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The change in reactor power from 70 percent to 40 percent was not as a result of the failed breaker, rather Browns Ferry Unit 1's change in reactor power was due to a scheduled reactor shutdown which was in progress. In regards to the Unit 2 loss of channel 'A' RPS, this was not a specified system actuation. The actuation of the 'A' and 'B' diesel generators were the specified system actuation. Although the 'A' and 'B' diesels are common to both Units 1 and 2, only Unit 1 credits these specific diesel generators for accident mitigation. As such, this event is only reportable from Unit 1. Unit 2 did not experience a specified system actuation. |
| ENS 57245 | 25 July 2024 20:48:00 | Browns Ferry | NRC Region 2 | The following information was provided by the licensee via phone and email: Tennessee Valley Authority (TVA) completed an engineering evaluation for a Fike Metal Products 16-inch rupture disc, part number 16-CPV-C, which had failed in March 2024 during an event previously reported to the NRC as Event Notification 57036 and Licensee Event Report 260/2024-002-00. The evaluation determined that the failure of the rupture disc constituted a failure to comply by a basic component which resulted in a substantial safety hazard. The rupture disc was procured as a non-safety related item from Fike Corporation and commercially dedicated by Paragon Energy Solutions. The disc was supplied to TVA in a satisfactory condition meeting all acceptance criteria. During a routine flowrate surveillance test, the high-pressure coolant injection (HPCI) inner rupture disc developed a hole which caused the Unit 2 HPCI turbine to trip. This resulted in (Browns Ferry Unit 2) entering Technical Specification (TS) Limiting Condition for Operation (LCO) 3.5.1 Condition `C', which is a 14-day shutdown LCO. Per HPCI system design criteria, turbine casing protection disc rupture pressure shall be at 175 psig plus 1 or minus 10 psig and the rupture discs shall be sized for a flow capacity of 600,000 pounds per hour at 200 psig, minimum. The failed HPCI inner rupture disc did not experience pressures above 45 psig since being installed; therefore, the HPCI turbine inner rupture disc did not meet its technical requirements. On July 23, 2024, the Browns Ferry Nuclear Plant Site Vice President was notified of the requirement to report this event under 10 CFR 21.21. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i). A written notification in accordance with 10 CFR 21.21(d)(3)(ii) will be provided within 30 days. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Both Fike Corporation and Paragon Energy Solutions have been informed of the HPCI inner rupture disc not meeting technical requirements. Known potentially affected plants include Browns Ferry Units 1, 2, and 3.
The following information was provided by the licensee via phone and email: The purpose of this notification is to retract a previous event notification, EN 57245, reported on 7/25/24. Continued evaluation has concluded that the failure of the disc was not the result of a failure to comply by a basic component, therefore, the NRC non-emergency 10 CFR 21.21 (d) report was not required and the NRC EN 57245 can be retracted. The licensee has notified the NRC Resident Inspector. Notified R2DO (Masters) and Part 21/50.55 Reactors group (Email). | |
| ENS 57090 | 25 April 2024 02:22:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 4/24/2024 at 2215 CDT, Browns Ferry Unit 1 experienced an automatic reactor scram. The cause of the scram is currently under investigation. The main steam isolation valves (MSIVs) remain open with the main turbine bypass valves controlling reactor pressure. The reactor feedwater pumps are in service to control reactor water level. Primary containment isolation systems (PCIS) Groups 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all components actuated as required. Following the reactor scram, due to reactor water level reaching minus 45 inches, both high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) initiation signals were received, and both initiated as designed. All safety systems operated as expected. This event requires a 4-hour report per 10 CFR 50.72(b)(2)(iv)(A), `Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. This event requires a 4-hour report per 10 CFR 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. 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): 1) Reactor protection system (RPS) including: reactor scram or reactor trip. 2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). 4) ECCS for boiling water reactors (BWRs) - high-pressure coolant injection (HPCI). 5) BWR reactor core isolation cooling system (RCIC). All safety systems operated as expected. At no time was public health and safety at risk. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Units 2 and 3 were not affected. |
| ENS 57070 | 10 April 2024 11:23:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email and phone call: A non-licensed employee supervisor had a confirmed positive during a random fitness-for-duty test. The employees access to the plant has been terminated. The NRC Senior Resident Inspector has been notified. |
| ENS 57036 | 19 March 2024 18:19:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via fax or email: While performing a planned high pressure coolant injection (HPCI) system surveillance, an isolation signal was received based upon an exhaust rupture disc high pressure signal. This resulted in an unplanned inoperability of the HPCI system. All systems responded as expected, and the event is under investigation. No other systems were affected by this condition. This event is reportable as an 8-hour non-emergency notification under 10CFR50.72(b)(3)(v) as HPCI is a single train safety system. There was no impact to plant personnel or the public as a result of this condition. The NRC resident has been notified of this condition. |
| ENS 56990 | 24 February 2024 09:27:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: At 0219 CST on February 24, 2024, Browns Ferry Unit 3 was shut down in a refueling outage, while closing 4 kV shutdown board breaker 3EB-9, the 4 kV shutdown board normal feeder breaker tripped open resulting in a valid 4 kV bus under-voltage condition. Due to the under-voltage condition, the 3B emergency diesel generator (EDG) auto started and tied to the board. The cause of the breaker tripping open is unknown and an investigation is in progress. All systems responded as expected for the loss of voltage. This event requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No other safety related equipment was affected. The 3B EDG continues to supply the shutdown board pending further investigation. |
| ENS 56941 | 1 February 2024 15:32:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On February 1, 2024, a contract worker was transported offsite for medical treatment due to a work-related injury that required the individual to be admitted to the hospital. The individual was free-released from the site prior to transport. The injury and hospitalization were reported by the contract worker's employer to OSHA per 29 CFR 1904.39(a)(2). Based upon that notification to another government agency, Tennessee Valley Authority is reporting this per 10 CFR 50.72(b)(2)(xi). The NRC Senior Resident Inspector has been notified of this event.
The following information was provided by the licensee via email: The purpose of this notification is to retract a previous Event Notification, EN 56941, reported on 02/01/2024. On 02/01/2024, at 15:32 EST, Browns Ferry Nuclear Plant (BFN) made an Event Notification 56941 notifying the NRC of a notification to another government agency. During further review of NRC reporting guidance, BFN has concluded that the contract worker's employer report to OSHA was below the reporting threshold outlined in NUREG 1022, Revision 3. The NRC Resident Inspector has been notified. |
| ENS 56531 | 20 May 2023 07:45:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: On 5/20/2023 at 0315 CDT, Browns Ferry Unit 1 was at 80 percent reactor power performing, 'Turbine control valve fast closure turbine trip and RPT (recirculation pump trip) initiate logic testing'. During performance of this test, Unit 1 received a full reactor scram. An investigation is in progress to determine the cause of the scram. All systems responded as expected, and Unit 1 is stable at zero percent power in mode 3. All control rods fully inserted into the core. Main steam isolation valves remained open with main turbine bypass valves controlling pressure. Reactor feedwater pumps remained in service to control reactor water level. Primary containment isolation signals groups 2, 3, 6, and 8 were received with expected system actuations. This event is reportable per 10 CFR 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'. The event is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. The NRC Resident has been notified. |
| ENS 56505 | 5 May 2023 16:00:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 05/04/2023 at 2034 CDT, a Browns Ferry Nuclear Plant non-licensed employee supervisor had a confirmed positive drug test identified during random fitness-for-duty medical testing. Employee's unescorted access has been suspended. A review of the employee's work has been completed. The (NRC) Resident Inspector has been notified. |
| ENS 56411 | 15 March 2023 04:27:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 2257 (CDT) on 3/14/2023 during the 2R22 refueling outage on Browns Ferry Nuclear Plant Unit 2, it was determined there was RCS boundary leakage from five of eight sensing lines that pass through containment penetrations X-30 and X-34 that did not meet the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code. The condition will be resolved prior to plant startup. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: The purpose of this notification is to retract a previous Event Notification, EN 56411 reported on 3/14/23. Following the initial notification, further analysis of the condition was performed. It was determined that the leaking pipe weld was ASME Section XI Code Class 2 piping which falls under the requirements of ASME Section XI Subsection IWC and not Subsection IWB. Therefore, this condition does not represent a serious degradation of the nuclear power plant, including its principle safety barriers. Based upon the above, the leaks identified on the ASME Section XI Code Class 2 equivalent Main Steam sense lines are not reportable under 10 CFR 50.72(b)(3)(ii). Therefore, the NRC non-emergency 10 CFR 50.72(b)(3)(ii) report was not required and the NRC report 56411 can be retracted and no Licensee Event Report under 10 CFR 50.73(a)(2)(ii) is required to be submitted. Notified R2DO (Miller) |
| ENS 56385 | 2 March 2023 17:52:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 1312 CST on March 2, 2023, Browns Ferry Nuclear Plant Units 1, 2, and 3 initiated voluntary communication to the state of Alabama and local officials as part of the Nuclear Energy Institute (NEI) Groundwater Protection Initiative (GPI), after receiving analysis results for leakage from a demineralized water storage tank that contained activity above the GPI voluntary communication threshold. All these results are significantly less than the limits established by the Nuclear Regulatory Commission (NRC) and Environmental Protection Agency (EPA) for effluents from the station. Further samples obtained of the water prior to entering the Tennessee River were less than detectable. The leakage source has been isolated and additional corrective actions are in progress. This condition did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56371 | 18 February 2023 11:25:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On February 17, 2023 during the planned U2R22 outage on Browns Ferry Nuclear Plant Unit 2, personnel entered the Unit 2 drywell for leak identification. Personnel discovered a cracked weld on the 2A recirculation pump discharge isolation valve drain line. At 0439 CST on February 18, 2023, following engineering evaluation, this drain line was determined to be 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 56321 | 24 January 2023 08:43:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 0121 CST on 01/24/2023, it was discovered that the Unit 1 High Pressure Coolant Injection System (HPCI) was inoperable; therefore, the condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), 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. 1-FCV-073-0006B, HPCI Steam Line Condensate Outboard Drain Valve, failed closed during normal plant configuration. This valve is normally open. The HPCI steam line is not being drained with the valve in the current position. The Unit 1 Nuclear Unit Senior Operator entered Unit 1 Technical Specifications LCO 3.5.1 Condition C with required actions C.1 to immediately verify by administrative means that the Reactor Core Isolation Cooling (RCIC) system is operable and C.2 to restore HPCI to operable status in 14 days. RCIC has been verified operable by administrative means. There was no impact to the safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56257 | 3 December 2022 13:05:00 | Browns Ferry | NRC Region 2 | GE-4 | 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 56168 | 18 October 2022 22:08:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 10/18/2022 at 1440 CDT, Browns Ferry Unit 3 declared both trains of standby liquid control (SLC) inoperable due to acceptance criteria failure of 3-SI-3.1.7.6, 'Standby Liquid Control System ATWS Equivalency Calculation for Newly Established Pump Flow Rate.' The purpose of this surveillance is to ensure the anticipated transient without scram (ATWS) calculation criteria is met after each pump flow test. Chemistry performed the surveillance following pump flow testing and the requirement for equivalency calculation failed low with a result of less than 1.0. CR 1810303 documents this condition in the corrective action program. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(A), 10 CFR 50.72(b)(3)(v)(C), and 10 CFR 50.72(b)(3)(v)(D). This condition is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(v)(A),10 CFR 50.73(a)(2)(v)(C), and 10 CFR 50.73(a)(2)(v)(D). The NRC Resident Inspector has been notified of this event. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officer's report guidance: The plant entered an 8 hour limiting condition for operation based on the above. The condition was resolved at 2053 CDT when the system was restored to normal operation. |
| ENS 55992 | 12 July 2022 17:25:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via fax or email: At 0917 CDT on 7/12/2022, during the performance of U1 (Unit 1) High Pressure Coolant Injection (HPCI) rated flow test, the 1-FCV-73-19 (HPCI governor valve) failed to operate as expected. This condition results in U1 HPCI being inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The Automatic Depressurization System (ADS) and Reactor Core Isolation Cooling (RCIC) system remain operable. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. U1 entered TS LCO 3.5.1 Condition C, 14-day Shutdown LCO (Limiting Condition for Operation), due to the HPCI inoperability. |
| ENS 55818 | 2 April 2022 15:10:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via fax: At 1345 CDT, Browns Ferry declared a Notification of Unusual Event due to a fire at the 3B Reactor Feedwater Pump within the Turbine Building which was not extinguished within 15 minutes. Subsequently, the fire was extinguished at 1402 CDT. Unit 3 remains in Mode 1 at approximately 9.5 percent rated thermal power (RTP). Unit 1 and 2 remain at 100 percent RTP and unaffected. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The fire began at 1332 CDT. It is believed that the fire was in the oil system of the Feedwater Pump. The fire was extinguished by the on-site fire brigade. No off-site assistance was requested. The Unusual Event was declared under Emergency Action Level HU-4. The licensee notified the NRC Resident Inspector and required State and local government agencies. Unit 3 is currently stable. Notified DHS-SWO, FEMA Ops Center, and CISA Central Watch Officer, and FEMA NWC, DHS NRCC THD Desk, and DHS NuclearSSA via email.
The Notification of Unusual Event was exited at 1544 CDT. Notified R2DO (Miller), IR-MOC (Kennedy), NRR-EO (Felts), DHS-SWO, FEMA Ops Center, and CISA Central Watch Officer. |
| ENS 55706 | 16 January 2022 06:41:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 1/15/2022 at 2320 (EST) during the planned F108 outage on Browns Ferry Nuclear Plant Unit 1, personnel entered the Unit 1 Drywell for leak identification. Personnel discovered a through-wall piping leak on a 3/4 inch test line upstream of the test valve. This 3/4 inch test line is located on the RHR Shutdown Cooling & RHR Return Line to the reactor vessel and 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 has been notified. |
| ENS 55660 | 16 December 2021 14:57:00 | Browns Ferry | NRC Region 2 | GE-4 |
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 Reactor Protection System (RPS). On October 20, 2021, at approximately 0705 hours Central Daylight Time (CDT), Browns Ferry, Unit 1, 1B RPS bus unexpectedly lost power. The loss of the bus resulted in a half scram, automatic Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolations, and Trains A, B, and C SBGT (Stand-By Gas Treatment) and A CREV (Control Room Emergency Ventilation system) started. All systems responded as expected. At 0720 hours CDT, the bus was placed on the alternate power supply and the half scram and PCIS isolations were reset. 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 cause of the RPS bus loss was a trip of the underfrequency relay due to drift of the relay setpoint. The relay was replaced and 1B RPS bus was returned to the normal power supply on October 21, 2021, at 0510 hours CDT. 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 1729592. The NRC Resident Inspector has been notified of this event. |
| ENS 55287 | 1 June 2021 17:46:00 | Browns Ferry | NRC Region 2 | GE-4 | 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 55200 | 20 April 2021 15:36:00 | Browns Ferry | NRC Region 2 | GE-4 | A non-licensed, employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. |
| ENS 54983 | 5 November 2020 06:32:00 | Browns Ferry | NRC Region 2 | GE-4 | At 2150 CST on 11/04/2020, it was discovered that Unit 1 High Pressure Coolant Injection System (HPCI) was INOPERABLE; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), 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. During performance of 1-SR-3.5.1.7, HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at Rated Reactor Pressure, Unit 1 HPCI was manually tripped by the control room operator due to local report of excessive shaking of the cooling water supply from the booster pump line. There was no impact to the safety of the public or plant personnel. The NRC Resident Inspector has been notified. CR 1650042 documents this condition in the Corrective Action Program. The Unit is in a 14-day LCO 3.5.1(c). The RCIC System is operable.
ENS Event number 54983, made on 11/05/2020 is being retracted. NRC notification 54983 was made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 were met when Unit 1 HPCI was manually tripped by the control room operator due to a local report for excessive shaking of the cooling water supply from the booster pump line. A subsequent engineering evaluation concluded on 11/06/2020 there was reasonable assurance of operability with no additional intrusive maintenance performed and that the condition was bounded by a previous evaluation documented in (Condition Report) CR 1347736. As such, the circumstances discussed in the report did not result in any condition that at the time of discovery could have prevented the fulfillment of the safety function of structures of the system that are needed to mitigate the consequences of an accident. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(v). TVA's evaluation of this event is documented in the corrective action program. The licensee has notified the NRC Resident Inspector. Notified R2DO (Miller). |
| ENS 54932 | 5 October 2020 14:25:00 | Browns Ferry | NRC Region 2 | This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On August 6, 2020, at approximately 1749 CDT, Browns Ferry Nuclear Plant (BFN), Unit 2 experienced a loss of Reactor Protection System (RPS) Bus 2A. Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolated in response to this event. The PCIS isolations caused the initiation of Standby Gas Treatment (SBGT) trains A, B, and C, and Control Room Emergency Ventilation (CREV) subsystem A. Unit 2 declared RCS leakage detection instrumentation inoperable and entered TS LCO 3.4.5 condition A, B, and D with required action D.1 to enter LCO 3.0.3 immediately. Unit 2 entered TS LCO 3.0.3 with required actions to be in Mode 2 within 10 hours, Mode 3 within 13 hours, and Mode 4 within 37 hours. Upon investigation, it was discovered that an age-related overheating condition resulted in the failure of the 2A RPS Motor Generator (MG) set, causing the feeder beaker from the 2A 480v Remote Motor-Operated Valve distribution board to trip. On August 6, 2020, at approximately 1808 CDT, Operations personnel commenced restoration of Unit 2 to normal after transferring 2A RPS to its alternate power supply. The 2A RPS MG Set drive motor was replaced on August 24, 2020. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel (RV) Low Water Level or Drywell High Pressure. Plant conditions which initiate PCIS Group 3 actuations are RV Low Water Level or Reactor Water Cleanup Area High Temperature. Plant conditions which initiate PCIS Group 6 actuations are RV Low Water Level, High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation. Plant conditions which initiate PCIS Group 8 actuations are Reactor Vessel (RV) Low Water Level or Drywell High Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. All affected safety systems responded as expected. 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 1628707. The NRC Resident Inspector has been notified of this event. | |
| ENS 54931 | 5 October 2020 13:51:00 | Browns Ferry | NRC Region 2 | This 60-day telephone notification is being made under 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 an emergency service water system component that does not normally run and which provides an ultimate heat sink. On August 6, 2020, at approximately 0128 CDT, the A3 Emergency Equipment Cooling Water (EECW) pump received an auto-start signal while performing Post-Maintenance Testing (PMT) on the 3C Core Spray pump. Normally, the involved EECW pump would be started prior to testing to prevent an auto-start; however, in this case the pump was not running prior to the test. When the 3C Core Spray pump breaker was closed while in the test position, an unanticipated actuation of the A3 EECW pump occurred. Work was stopped and the workers reported to the Control Room to evaluate the condition. Based on a review of this event, individuals involved were coached on understanding system response prior to performing work. The A3 EECW pump responded in accordance with the plant design. No other plant equipment was affected during this event. There were no safety consequences or impacts on the health and safety of the public. The event was entered into TVA's corrective action program for evaluation and resolution. Reference corrective action document CR 1628479. The NRC Resident Inspector has been notified of this event. | |
| ENS 54795 | 21 July 2020 08:58:00 | Browns Ferry | NRC Region 2 | 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 | Browns Ferry | NRC Region 2 | 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 54709 | 13 May 2020 13:10:00 | Browns Ferry | NRC Region 2 | 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 March 16, 2020, at approximately 0102 CDT, Browns Ferry Nuclear Plant (BFN), Unit 3 received motor trip-out alarms and diagnosed Group 2 and 3 Primary Containment Isolation System (PCIS) Isolations, 3C Residual Heat Removal (RHR) Pump tripping and Reactor Water Cleanup (RWCU) system isolating. All affected safety systems responded as expected. BFN, Unit 3, was nearing the end of the U3R19 refueling outage at the time of the event, and was still dependent on the Shutdown Cooling (SDC) system. 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. At the time of the event, these conditions did not exist: therefore, the PCIS actuation was invalid. The event was determined to have been caused by clearance restoration activities in an unprotected control panel. A fuse re-installation inadvertently created a fault condition between two different plant 120 VAC power sources when the fuse holder's lower spring clip contacted a different fuse. This was a result of age-related degradation of the fuse holder, its close proximity to other fuses, and the lack of insulating isolation barriers between fuses. 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 1594925. The NRC Resident Inspector has been notified of this event. | |
| ENS 54697 | 6 May 2020 14:11:00 | Browns Ferry | NRC Region 2 | 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 March 8, 2020, at approximately 2021 CDT, Browns Ferry Nuclear Plant Unit 2 experienced an unexpected loss of the 2A Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and the initiation of Standby Gas Treatment Trains A and B, and Control Room Emergency Ventilation System Train A. All affected safety 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. The RPS MG Set trip was believed to have been caused by an intermittent short across a spike suppressor, which led to a loss of generator output signal to a voltage regulator. The affected components have been replaced. 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 1593265. The NRC Resident Inspector has been notified of this event. | |
| ENS 54559 | 3 March 2020 12:18:00 | Browns Ferry | NRC Region 2 | At 0416 CST on March 3, 2020, with Browns Ferry Unit 3 in Mode 5, there was a partial loss of 161kV power due to a trip and re-closure of the feeder breaker supplying the Common Station Service Transformer (CSST) A. Unit 3 lost power to the 1A Start Bus and the 3A Unit Board, which supplies power to the 3EA and 3EB 4kV Shutdown Boards. The loss of the 4kV Shutdown Boards resulted in an auto actuation of the 3A and 3B Emergency Diesel Generators (EDG). The 3C EDG and 3EC 4kV Shutdown Board were removed from service for maintenance at the time of the loss of power, and the 3ED 4kV Shutdown Board continued to be powered from the 3B Unit Board. All systems responded as expected for the loss of power. This event requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The partial loss of power was caused by a storm in the area. Offsite power has been restored to the CSST and the EDGs have been secured. | |
| ENS 54382 | 12 November 2019 10:32:00 | Browns Ferry | NRC Region 2 | On November 12, 2019, the Central Emergency Control Center (CECC) was removed from service for a planned facility upgrade project. The CECC is a common Emergency Operations Facility (EOF) for the TVA Nuclear sites (Browns Ferry / Sequoyah / Watts Bar). The duration of the upgrade project is approximately 75 days. If an emergency is declared requiring CECC activation during this period, an alternate CECC will be used. During this period, the alternate CECC will be staffed and activated using existing emergency procedures. This is an eight-hour, non-emergency notification for a Loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the CECC will be unavailable for more than 72 hours. The Emergency Response Organization has been notified that the CECC will be unavailable during the upgrade project and to report to the alternate CECC in the event of an emergency. There is no impact on the health and safety of the public or plant employees. The NRC Resident Inspector has been notified.
The event information was updated to indicate that the event occurred at 0700 EST. The NRC Resident Inspector has been notified. Notified the R2DO(Musser).
The CECC facility upgrade project is sufficiently complete such that the CECC was returned to a functional status at 1350 CST on January 31, 2020. The NRC Resident Inspector has been notified. Notified R2DO (Baptist). | |
| ENS 54341 | 21 October 2019 15:52:00 | Browns Ferry | NRC Region 2 | 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 December 29, 2018, at approximately 0220 Central Standard Time (CST), Browns Ferry Nuclear Plant (BFN), Unit 3 experienced an unexpected loss of power to the 3A Reactor Protection System (RPS) Bus due to the trip of the 3A RPS motor generator (MG) set. This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected. This event is being reported as a late 60 day non-emergency notification. This missed notification was identified on August 23, 2019. 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 cause of the trip of the RPS MG Set was a failure of the motor winding insulation of all three phases. 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 Reports 1478564 and 1543534. The NRC Resident Inspector has been notified of this event. | |
| ENS 54332 | 16 October 2019 10:22:00 | Browns Ferry | NRC Region 2 | 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 August 20, 2019, at approximately 1133 hours Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN), Unit 2 experienced an unexpected loss of the 2A Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety 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. The cause of the RPS MG Set trip was dirty potentiometer windings on an Over Voltage Relay. The dirt prevented the potentiometer's wiper from contacting its windings, resulting in erratic setpoint values. 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 Reports 1542603, 1542608, and 1542569. The NRC Resident Inspector has been notified of this event. | |
| ENS 54302 | 1 October 2019 07:05:00 | Browns Ferry | NRC Region 2 | On 10/1/2019, at 0307 CDT, Unit 2 was conducting a normal reactor startup and received a valid Reactor Protection System (RPS) scram. The reactor was critical in MODE 2 at the Point of Adding Heat. Operators began withdrawing Source Range Monitor (SRM) Instrumentation per procedure. When the operator depressed the SRM Drive Out pushbutton to withdraw the last two SRMs (C and D), an unexpected full Reactor Scram was received. Annunciator indication in the Main Control Room indicated a Neutron Monitoring Scram. The Intermediate Range Monitors (IRM) D, E, F, H and G all indicated Upscale High High. There were no Emergency Core Cooling System (ECCS) or Containment Isolation System actuations. All other systems functioned as designed. The cause of the Reactor Scram is still under investigation. This event requires a 4-hour report per 10 CFR 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.' This event also 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), (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.' The NRC Resident Inspector has been notified. | |
| ENS 54300 | 30 September 2019 10:28:00 | Browns Ferry | NRC Region 2 | 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 July 31, 2019, at approximately 1650 hours Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN), Unit 1 experienced a Primary Containment Isolation System (PCIS) Group 6 isolation during performance of surveillance procedure 1-SR-3.3.6.2.3(A), Reactor/Refueling Zone Ventilation Radiation Monitor 1-RM-90-140/142 Calibration and Functional Test. The Group 6 isolation caused the initiation of Standby Gas Treatment (SBGT) Trains A, B, and C, and Control Room Emergency Ventilation (CREV) subsystem B. Unit 1 H2O2 Analyzer and Drywell Radiation Monitor CAM, 1-RM-90-256, were declared Inoperable and Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.5 Condition B was entered. All affected safety systems responded as expected. Plant conditions which initiate PCIS Group 6 actuations are Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. This condition was the result of two cleared fuses in the alarm logic. The apparent cause is a ground fault on the A6 Open Drain Input/Output Module. 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 Acton Program as Condition Report 1537358. The NRC Resident Inspector has been notified of this event. | |
| ENS 54278 | 13 September 2019 12:23:00 | Browns Ferry | NRC Region 2 | EN Revision Text: EMERGENCY OPERATING FACILITY UNAVAILABLE DUE TO ACCESS ISSUES This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. A condition impacting access to the Emergency Operating Facility, Central Emergency Control Center (CECC), located in the TVA Chattanooga Office Complex occurred on September 13, 2019 at 0427 CDT. Fire suppression capabilities for the TVA Chattanooga Office Complex are currently impacted by a water main failure rendering access to the facility unsafe for personnel. If an emergency is declared requiring CECC activation during this period, other emergency response centers will be activated and staffed using existing emergency planning procedures and have the capability to perform the functions normally performed by the CECC. This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the condition affects the functionality of an emergency response facility. The condition does not affect the health and safety of the public. The NRC Resident Inspector has been notified.
Water lines impacting the Chattanooga Office Complex were repaired, and as of time 0734 EDT on 9/16/19, the CECC was returned to a functional status. The NRC Resident Inspector has been notified. Notified R2DO (Ehrhardt). | |
| ENS 54266 | 11 September 2019 03:10:00 | Browns Ferry | NRC Region 2 | A lightning strike occurred at approximately 1502 CDT on 09/10/2019, and a resulting power surge damaged some of the security door card reader system equipment. However, this did not affect access to plant areas for personnel who were already within protected area. At 1830 on 09/10/2019, it was discovered that some of the oncoming night shift personnel could not access particular areas that required the use of security card readers. Extent of condition check at 1934 on 09/10/2019 determined that access to 1A and 3A Electric Board Rooms, which contain remote shutdown panels and Fire Safe Shutdown equipment. was prohibited for the night shift personnel. This condition is reportable under 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. Access was restored to all plant areas at 2106 on 9/10/2019. No plant events occurred during the time frame that the 1A & 3A Electric Board Rooms inaccessible that would have required access to these areas. The NRC Resident Inspector has been notified. | |
| ENS 54162 | 12 July 2019 22:50:00 | Browns Ferry | NRC Region 2 | 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 53959 | 26 March 2019 16:08:00 | Browns Ferry | NRC Region 2 | On 3/26/2019 at 1030 CDT Engineering evaluation determined that Traversing lncore Probe (TIP) System test results related to Leak Rate Testing of 2-CKV-76-653, TIP Purge Header Check Valve, during the Unit 2 Refueling Outage resulted in a reportable condition. On 3/24/2019 at 1558 CDT, Leak Rate Testing identified a (local leak rate test) LLRT failure of 2-CKV-76-653. The gross leakage Leak Rate value exceeded the Technical Specification allowable value for Type C valves of less than 0.6 (allowable leakage) La. 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. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The short-term corrective actions include repairing the valve such that it passes the test. The valve needs to be repaired before the unit can change modes. | |
| ENS 53942 | 17 March 2019 14:10:00 | Browns Ferry | NRC Region 2 | EN Revision Text: HIGH PRESSURE COOLANT INJECTION SYSTEM DECLARED INOPERABLE At 0735 CDT on March 17, 2019, the High Pressure Coolant Injection (HPCI) system was isolated due to a water-side leak from the HPCI Gland Seal Condenser. Unit 3 declared the HPCI system Inoperable and entered Technical Specification LCO 3.5.1 Condition C with required actions to verify the Reactor Core Isolation Cooling system is Operable, and to restore the HPCI system to Operable status within 14 days. All other Unit 3 Emergency Core Cooling Systems (ECCS) remain Operable. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(V)(D), '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.' This is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(V)(D). 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 Event Number 53942, made on March 17, 2019, is being retracted. NRC Notification 53942 was made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 (b)(3)(v)(D) were met when the licensee discovered an event, 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. At 0735 CDT, on March 17, 2019, during the performance of a routine surveillance, a momentary pressure transient of 844 psig from the Feedwater system was introduced into the High Pressure Coolant Injection (HPCI) system discharge and suction piping that ruptured the seal on the gland seal condenser and flooded the U3 HPCI Room. Unit 3 HPCI was declared inoperable due to isolation of the waterside of the HPCl system. On April 11, 2019, a Past Operability Evaluation was completed which determined that the HPCI System remained operable. The evaluation of the potential pressure transient and room flooding concluded that the HPCI System could have performed its specified safety function of vessel injection throughout the time that the gland seal was ruptured. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(v)(D). TVA's evaluation of this event is documented in the Corrective Action Program in Condition Report 149973. The licensee has notified the NRC Resident Inspector. Notified R2DO (Ehrhardt). | |
| ENS 53923 | 10 March 2019 04:38:00 | Browns Ferry | NRC Region 2 | At 2259 CST on 3/9/2019, Browns Ferry Unit-3 received an automatic SCRAM on Main Generator Breaker Failure and Turbine Load Reject. Unit-3 declared a Notification of Unusual Event SU1 for loss of offsite AC power to Unit-3 specific 4kV Shutdown Boards for greater than 15 minutes. Primary Containment Isolation Systems (PCIS) Groups 1, 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required. Main steam relief valves lifted on the initial transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated on low reactor water level. HPCI remains in service for reactor level and pressure control. RCIC is not in service at this time, the station is investigating low flow from the pump. All four Unit-3 Diesel Generators started and loaded as expected. Residual Heat Removal System is in service for suppression pool cooling. 4kV Station Unit Boards have been restored from the 161kV system. Actions are in progress to restore 4kV Shutdown Boards to offsite power. This event is reportable within 1 hour in accordance with 10 CFR 50.72(a)(1)(i) for declaration of the Licensees Emergency Plan. Complete as documented on EN 53922. This event requires a 4 hour report per 10 CFR 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.' This event also 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), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs), (4) ECCS (Emergency Core Cooling System) for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system, (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system, and (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified. As of the event report, the MSIVs were opened and decay heat was being removed via the bypass valves to the condenser. | |
| ENS 53922 | 10 March 2019 00:48:00 | Browns Ferry | NRC Region 2 | At 0012 EST on 3/10/2019, Browns Ferry Unit-3 declared an Unusual Event due to a spurious trip of the generator breaker, resulting in a loss of AC power to the 4 kV shutdown boards greater than 15 minutes. All diesel generators started and loaded to supply onsite power. The reactor auto-scrammed, with all rods fully inserting. The Main Steam Isolation Valves opened and shutdown cooling was being conducted via the condenser. The licensee will exit the emergency declaration once offsite power is restored. There is no estimated restart date. Browns Ferry Unit 1 remains in Mode-1 (100%), Unit 2 remains in Mode-5 for a refueling outage. The NRC Resident Inspector has been notified. This event is related to EN 53923. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
At 1310 CDT, Browns Ferry Unit-3 exited the Unusual Event when 161 kV lines were made available. The licensee is executing procedures for securing the diesel generators while alternate offsite power methods are utilized. Switchyard damage evaluation is in progress. The licensee will notify the NRC Resident Inspector. Notified R2DO (Desai), R2RA (Haney), DNRR (Nieh), NRR EO (Miller), and IRD (Grant). Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email). | |
| ENS 53918 | 8 March 2019 15:51:00 | Browns Ferry | NRC Region 2 | Browns Ferry Nuclear Plant (BFN) is notifying state and local agencies of the presence of an oil sheen in the cold water channel. Water from the cold water channel was running into a tunnel that connects to the waters of the US. BFN Procedure RWI-007, Spill Prevention Control and Countermeasure Plan requires the National Response Center as well as other state and local agencies be notified of any oil sheen on the water. This oil spill is reportable to the EPA (National Response Center) under 40 CFR 112. The notification was made to the National Response Center at 1113 CST under notification number 1239580. The Alabama Emergency Management Agency (AEMA) and Alabama Department of Environmental Management (ADEM) were notified at 1120 CST. This event is reportable as a 4-hour Non-Emergency Notification report in accordance with 10 CFR 50.72(b)(2)(xi) 'Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.' The licensee has notified the NRC Resident Inspector. The oil is believed to come from the number one cooling tower basin due to heavy rainfall. | |
| ENS 53750 | 22 November 2018 03:56:00 | Browns Ferry | NRC Region 2 | EN Revision Text: HPCI UNEXPECTEDLY TRANSFERRED TO ALTERNATE SUCTION SOURCE DURING TESTING At 2125 (CST) on 11/21/2018, it was discovered that U1 High Pressure Coolant Injection System (HPCI) was inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), 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. During performance of a routine surveillance, HPCI automatically transferred from its normal suction source to the alternate suction source. The control room operator then manually tripped the HPCI turbine. HPCI was already inoperable in accordance with Technical Specifications (TS) Limiting Condition for Operability (LCO) 3.5.1, ECCS Operating, Condition C during performance of the surveillance. However, this condition was not expected nor induced by the testing. There was no impact to the safety of the public or plant personnel. The NRC Resident Inspector has been notified. CR 1469109 documents this condition in the Corrective Action Program.
ENS Event Number 53750, made on November 22, 2018, is being retracted. NRC notification 53750 was made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72(b)(3)(v)(D) were met when the licensee discovered 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. During performance of a routine surveillance, the High Pressure Coolant Injection (HPCI) System automatically transferred from its normal suction source to the alternate suction source. As a result, Unit 1 HPCI was declared inoperable. On December 20, 2018, a Past Operability Evaluation was completed which determined that the HPCI System remained operable. The evaluation determined that the HPCI System could have performed its specified safety function of vessel injection throughout the time that the suction path was aligned to the torus. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(v)(D). TVA's evaluation of this event is documented in the Corrective Action Program in Condition Report 1469109. The licensee has notified the NRC Resident Inspector. Notified R2DO (Desai). | |
| ENS 53678 | 21 October 2018 06:19:00 | Browns Ferry | NRC Region 2 | 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 53661 | 11 October 2018 15:37:00 | Browns Ferry | NRC Region 2 | 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 August 16, 2018, at approximately 1736 CDT, Browns Ferry Nuclear Plant (BFN), Unit 2 experienced an unexpected loss of the 2B Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected with the exception of the Unit 1 Refuel Zone Supply Fan Outboard Isolation Damper, 1-FCO-64-5, that failed to indicate closed position. 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 cause of the RPS MG (Motor Generator) Set trip was a failed (shorted) operating coil associated with the 480 VAC motor starter inside the control box. 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 Reports 1440047 and 1440050. The NRC Resident Inspector has been notified of this event. | |
| ENS 53558 | 22 August 2018 19:35:00 | Browns Ferry | NRC Region 2 | On 08/22/2018 at 1803 hours CDT, Browns Ferry Nuclear Plant declared an Unusual Event per EAL HU4, a fire potentially degrading the level of safety of the plant. At 1748 CDT Unit 1 received a call reporting smoke coming from the 480V Condensate Demineralizer Panel 3 in the Unit 3 turbine building elevation 557'. At approximately 1803 (CDT), the incident commander on the scene confirmed a fire inside the panel and all three units entered 0-AOI-26-1, Fire Response. The board was subsequently de-energized by operations personnel and the fire was extinguished at 1806 CDT. SM (Shift Manager) exited EAL HU4 and all three units exited 0-AOI-26-1 at 1840 CDT. Fire operations remain on scene to monitor. A team is being assembled for damage assessment and recovery. The fire did not affect any safety systems, no plant transients resulted, and no injuries were reported. This event is reportable within 1 hour IAW 10 CFR 50.72(a)(1)(i). The NRC Resident inspector has been notified. Notified DHS SWO, FEMA Ops, DHS NICC, FEMA NWC (email) and NuclearSSA (email). | |
| ENS 53497 | 9 July 2018 17:01:00 | Browns Ferry | NRC Region 2 | On 07/09/2018 at 1111 CDT, Browns Ferry Unit 1 Operators identified U1 High Pressure Cooling Injection system steam supply valves were isolated. After reviewing ICS (Integrated Computer System), Operations determined isolation occurred at 0958 CDT during performance of surveillance testing. The Browns Ferry Nuclear Plant Unit 1 High Pressure Coolant Injection (HPCI) system was declared inoperable at 0958 CDT due to an inadvertent isolation that occurred during testing. During performance of surveillance procedure 1-SR-3.3.6.1.2(3B) HPCI System Steam Supply Low Pressure Functional test, an erroneous signal was induced causing actuation of primary containment isolation system group IV (i.e., HPCI Isolation). 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). The erroneous signal was cleared and the HPCI isolation was reset. Upon reset of the isolation signal, the HPCI system was returned to available status. The HPCI system was unavailable for 2 hours and 55 minutes, however the HPCI system remains inoperable. There was no impact to the health and safety of the public or plant personnel as a result of this condition. The NRC Resident Inspector has been notified. A condition report has been entered into the Licensee's Corrective Action Program to capture this event. | |
| ENS 53462 | 20 June 2018 11:39:00 | Browns Ferry | NRC Region 2 | 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 53340 | 17 April 2018 12:02:00 | Browns Ferry | NRC Region 2 | GE-4 | At 0416 CDT on April 17, 2018, the High Pressure Coolant Injection System (HPCI) was isolated due to a water side leak from the gland seal condenser. Unit 1 HPCI remains inoperable pending repairs to the gland seal condenser. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(V)(D), '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.' This is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(V)(D). 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 53300 | 29 March 2018 22:28:00 | Browns Ferry | NRC Region 2 | GE-4 | 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. |