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| Entered date | Event description | |
|---|---|---|
| ENS 55706 | 16 January 2022 06:41:00 | 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 54559 | 3 March 2020 12:18:00 | 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 54302 | 1 October 2019 07:05:00 | 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 53959 | 26 March 2019 16:08:00 | 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 | 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 54341 | 21 October 2019 15:52:00 | 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 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 52707 | 25 April 2017 17:53:00 | Following engineering evaluation of tornado missile protection, the Emergency Diesel Generator (EDG) 7 day tank vent piping is subject to potential damage for EDGs D, 3A, 3B, 3C, and 3D. In the event that a tornado missile impact occurs on the aforementioned ventilation piping, there is a possibility that the vent lines could crimp. This could prevent the tanks from venting and would inhibit the transfer of fuel oil from the 7 day tank to the associated EDG. As a result, EDGs D, 3A, 3B, 3C, and 3D are inoperable for tornado missile protection and the Tech Spec actions cannot be met. These conditions are reportable IAW 10CFR50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. This issue is being addressed IAW EGM 15-002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance. The NRC Resident Inspector has been notified. |
| ENS 51850 | 6 April 2016 17:07:00 | At 1545 CDT on 04/06/16 Browns Ferry Unit 3 declared and exited the declaration of an unusual event due to a main steam line high high radiation condition. Power to Unit 3 was reduced to 91 percent power. The high radiation condition alarm cleared at 1526 CDT. Browns Ferry Unit 3 reported that the high radiation conditions were due to resin intrusion from the condensate demineralizers into the reactor and hydrogen water chemistry was a potential contributor to the event. The cause is still under investigation. The NRC Resident Inspector has been notified. State and Local notifications were made. Notified DHS SWO, FEMA Ops Center, NICC Watch Officer, FEMA NWC and Nuclear SSA (email).
At 1941 (CDT) BFN (Brown's Ferry) determined this notification to be potentially newsworthy due to receiving notification that counties (surrounding the plant) were alerted of this event. No plant conditions changed. The licensee may issue a press release. The licensee will notify the NRC Resident Inspector. Notified R2DO (Nease). |