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| Entered date | Event description | |
|---|---|---|
| ENS 51333 | 20 August 2015 18:51:00 | On August 20, 2015, at approximately 1032 CDT, during the Residual Heat Removal flow rate test, the 3ED 4kV Shutdown Board received a degraded voltage signal, which automatically started the 3D Emergency Diesel Generator (EDG). The EDG performed its safety function by automatically supplying power to the Shutdown Board. Troubleshooting on the degraded voltage signal is in progress. The remaining 4kV Shutdown Boards and EDGs were unaffected and remain operable. This constitutes an automatic actuation of the EDG and requires an 8-hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A), due to the valid actuation of the EDG, and a 60-day report in accordance with 10 CFR 50.73(a)(2)(iv)(A). The licensee has notified the NRC Resident Inspector. |
| ENS 45728 | 25 February 2010 23:37:00 | At 1840 on 2/25/10, Unit 2 RHR Division II became inoperable due to exceeding 260 degrees F on the injection line. The temperature is elevated due to leakage past the injection line check valve (2-CKV-74-68) and the inboard injection valve (2-FCV-74-67). The 260 degree F value is based on engineering calculations to ensure the injection line does not become steam voided. RHR Division I was already inoperable for scheduled maintenance when this condition occurred. Entered TS LCO 3.0.3 based on TS LCO 3.5.1 Condition H - Two or more Low pressure ECCS injection/spray subsystems inoperable for reasons other than Condition A. Action was taken to lower power at 1935 when the conditions requiring entry into LCO 3.0.3 could not be corrected. Tech Spec LCO 3.0.3 was exited at 2000 when RHR Loop II system pressure was raised using an alternate keep fill flow path to ensure that the injection line would remain filled. Power will be returned to 100%. This event requires a 4 hour report in accordance with 10 CFR 50.72(b)(2)(i). The licensee informed the NRC Resident Inspector. |
| ENS 45494 | 12 November 2009 22:36:00 | The HPCI (High Pressure Coolant Injection) system was declared inoperable after completion of a scheduled surveillance due to an excessive amount of water in the turbine exhaust line. The line was being drained in response to the alarm 'HPCI TURB EXH DRAIN POT LEVEL HIGH', indicating that there was a high level in a drain pot attached to the turbine exhaust line. An investigation is in progress to determine the source of the water in the turbine exhaust line. This event is reportable within 8 hours in accordance with 10CFR 50.72(b)(3)(v) as an event or condition that at the time of discovery could have prevented the fulfillment of a safety function. It also requires a 60 day written report in accordance with 10CFR 50.73(a)(2)(vii) The NRC Resident Inspector has been notified. SR number associated with this report: 91546. |
| ENS 45320 | 1 September 2009 21:01:00 | Unit 1 High Pressure Coolant Injection (HPCI) system was declared inoperable after the 250V RMOV BD breaker 11A1 for the ECCS DIV II Inverter tripped. The cause for the breaker trip is currently being investigated. This event is reportable within 8 hours in accordance with 10CFR 50.72(b)(3)(v) as an event or condition that at the time of discovery could have prevented the fulfillment of a safety function. It also requires a 60 day written report in accordance with 10CFR 50.73(a)(2)(vii). Unit 1 remains at 100% power. NRC Resident Inspector has been notified. |
| ENS 45290 | 24 August 2009 23:38:00 | On 8/24/09, at 18:50 Unit 3 was manually scrammed due to loss of 2 of the 3 Condensate Booster Pumps due to low pump suction pressure. The cause for the Condensate Booster Pump low suction pressures is unknown at this time, but is under investigation. After the reactor was scrammed manually, reactor water level lowered below the automatic scram set point (+2 inches) and below the automatic start for HPCI and RCIC (-45 inches). All expected Primary and Secondary Containment isolation valves operated as required, isolation groups 2,3,6 and 8 were actuated. Both reactor recirculation pumps tripped due to the low reactor water level. HPCI and RCIC actuated as expected to restore reactor water level. Reactor pressure control was maintained on the turbine bypass valves, and no Main Steam Relief Valves (MSRVs) were opened as a result of the transient. At this time the unit is stable in mode 3. Reactor water level is being controlled using one Reactor Feedwater pump, HPCI and RCIC have been returned to standby readiness. The 3B Reactor Recirculation Pump has been returned to service. Reactor pressure is being automatically maintained by the main turbine bypass valves. This event is reportable as a 4 hour non-emergency report due to 10CFR 50.72(b)(2)(iv)(A) and (B) (ECCS discharge to the reactor and Reactor Protection System (RPS) actuation) and as an 8 hour non-emergency report due to 10CFR50.72(b)(3)(iv)(A) (specified system actuations). All rods fully inserted on the SCRAM. The plant is in its normal shutdown lineup. The licensee notified the NRC Resident Inspector. |
| ENS 44860 | 18 February 2009 08:03:00 | At 0351 on 2/18/09, the Unit 1 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip due to a power load unbalance signal on the main generator. The cause of the power load unbalance signal was due to a generator neutral over voltage condition of which the cause is unknown and the investigation is continuing. All systems responded as expected to the turbine trip. One Safety Relief Valve (SRV) opened due to the reactor pressure transient, and then reactor pressure was automatically controlled by the Main Turbine Bypass valves. No Emergency Core Cooling System (ECCS), or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached, all expected containment isolation and initiation signals were received, and reactor water level is being automatically controlled by the feedwater system. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(iv)(B) for 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. All control rods fully inserted. The plant electrical system is in normal shut down alignment. No Diesel Generators started as a result of this event. There was no ECCS injection to the reactor vessel.
Review of available data indicates that no Main Steam safety/relief valves (MSRVs) opened in response to the Unit 1 reactor scram on 02/18/2009. There were no indications of an open MSRV on any discharge tailpipe thermocouple or acoustic monitor. Initial indications of the discharge tailpipe thermocouples for MSRV (1-PCV-1-30) did indicate a slight increase in temperature (approximately 36 degrees F) as reactor pressure decreased, which resulted in the initial assumption of an SRV opening. However, this behavior is a classical indication of slight main seat leakage and system engineering believes this seat leakage is what the post scram data indicates. Utilizing multiple reactor pressure instrumentation responses, the peak reactor pressure was determined to be approximately 1130 psig which is 15 psi below MSRV 1-PCV-1-30 setpoint. Additionally, the rise in tailpipe temperature did not coincide with the peak pressure but was after pressure had lowered. Based upon a thorough review of this data and a better understanding of the timing of the temperature rise, it is now believed that the MSRVs performed as designed during the reactor pressure transient event. The initial determination that concluded an MSRV opened will be further investigated within the corrective action program; reference PER 164114." The licensee has notified the NRC Resident Inspector. Notified R2 DO (Charlie Payne) |
| ENS 44854 | 16 February 2009 09:45:00 | At 0513 on 2/16/09, the Unit 2 reactor was manually scrammed in accordance with alarm response procedure 2-ARP-9-8A 'TURBINE TRIP TIMER INITIATED'. Other associated alarms and indications both locally and in the Main Control Room indicated a failure of the stator cooling water system. The exact cause of the failure is still being investigated. All systems responded as expected to the insertion of the manual scram. No ECCS injection was initiated or required, and all expected containment isolation and initiation signals were received. 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). All rods inserted fully into the reactor. The electrical power system is in a normal shut down configuration. Decay heat removal is through the main condenser via the turbine bypass valves. There is no impact on Units 1 and 3. The NRC resident inspector has been notified. |
| ENS 44252 | 1 June 2008 17:33:00 | A sample was taken of the lubrication/control oil from the Unit 1 High Pressure Coolant Injection (HPCI) System for moisture content analysis. The results of this analysis concluded that the moisture content in the oil exceeded acceptable levels. As a result of this condition the Browns Ferry Nuclear Unit 1 HPCI system was declared inoperable at 11:48 on 6/1/08. This event is reportable within 8 hours in accordance with 10 CFR 50.72 (b)(3)(v) as event or condition that at the time of discovery could have prevented the fulfillment of a safety function. It also requires a 60 day written report in accordance with 10 CFR 50.73 (a)(2)(vii). Unit 1 remains at 100% power. Unit 1 has entered Technical Specification LCO 3.5.1 and is performing the required actions. Troubleshooting of the moisture intrusion condition is in progress, and a corrective action plan is being developed. The NRC Resident inspector has been notified.
ENS Event Number 44252, made on June 1, 2008, is being retracted. NRC Notification 44252 was conservatively made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 were met when an oil sample indicated the Unit 1 High Pressure Coolant Injection (HPCI) System turbine oil system contained excessive amounts of water. An evaluation performed in response to this report concluded that the Unit 1 HPCI System was capable of performing its intended safety function with the turbine oil system containing more water than recommended in the TVA Lubrication Oil Analysis and Monitoring Program. TVA found through an engineering evaluation that the amount of water contained in the turbine oil system would not impact the HPCI operation during its mission time for the Design Basis Loss-of-Coolant Accident. 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 or systems that are needed to remove residual heat, or mitigate the consequences of an accident. Thus, there was no impact on nuclear safety. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(v)(B) or 10 CFR 50.72(b)(3)(v)(D). TVA's evaluation of this issue is documented in the corrective action program (PER 145517). The licensee has notified the NRC Resident Inspector. R2DO (Haag) notified. |
| ENS 43878 | 1 January 2008 01:58:00 | On 12/31/07 at 2140 the Unit 3 reactor scrammed due to turbine generator load reject signal on the Main Generator. The cause of the load reject signal is unknown and the investigation is continuing. All systems responded as expected to the load reject signal. Six Main Steam Relief valves (MSRVs) opened momentarily and then reclosed. Subsequently, reactor pressure was automatically controlled by the Main Turbine Bypass valves. No Emergency Core Cooling System (ECCS), nor Reactor Core Isolation Cooling (RCIC) reactor water level initiation setpoints were reached, and reactor water level is being automatically controlled by the Feedwater system. This report is being made as required by 10CFR 50.72(b)(2) due to the actuation of the Reactor protection System. Refer to BFN PER number 135878. All control rods fully inserted into the core, and all safety systems are operable. PCIS group isolations were received for groups 2, 3, 6, and 8. There were no grid abnormalities at the time of the load reject, and the event had no effect on Unit 1 or 2. The licensee notified the NRC Resident Inspector. |
| ENS 43659 | 22 September 2007 16:35:00 | As part of a planned outage for Browns Ferry Unit 3, initial drywell leak inspections were performed after shutdown (mode 3). This inspection identified a weld defect in Residual Heat Removal (RHR) piping. The defect was in a one inch test line near manually operated valve 3-74-638B. This is classified as pressure boundary leakage and the piping is rated as ASME code class 1. The leak rate was estimated by visual observation at less than 0.25 gpm. Investigation is continuing into the cause of the weld defect. Unit 1 and 2 remain at full power and are not affected by this event. This event is reportable within 8 hours under 10 CFR 50.72(b)(3)(ii)(A) and 10 CFR 50.73(a)(2)(ii)(A) as 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The licensee plans to continue to Mode 4 (Cold Shutdown) as required by Tech Specs for pressure boundary leakage. The NRC resident inspector has been notified. |
| ENS 40828 | 18 June 2004 16:31:00 | The following information was obtained by the Licensee via facsimile: A TVA employee at Browns Ferry Nuclear Plant was transported by ambulance to Athens-Limestone hospital, where he was later pronounced dead after becoming ill at work on Friday, June 18, 2004. The employee was part of a team of engineers inspecting a motor control unit inside the Unit 3 Reactor Building when he complained of shortness of breath. Emergency medical personnel in Browns Ferry's Fire Operations organization were called to respond and transferred the employee to the Athens-Limestone hospital ambulance service, which took him to the hospital in Athens. The engineering team was taking photographs and thermal images of the motor control unit while it was in operation. The medical emergency was not related to the operation of Unit 3 reactor. This event is being reported pursuant to 10CFR 50.72(b)(2)(xi) as '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. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials'. The employee was not contaminated. The Licensee has notified OSHA. The Licensee notified the NRC Resident Inspector. |