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| Entered date | Event description | |
|---|---|---|
| ENS 53922 | 10 March 2019 00:48:00 | 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 53923 | 10 March 2019 04:38:00 | 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 53247 | 7 March 2018 12:25:00 | The licensee declared an Unusual Event based on Emergency Action Level (EAL) 6.7.U and entry into the site Security Plan. All required actions or compensatory measures have been completed. The Notice of Unusual Event was terminated at 1142 CST. There was no impact to the operation of any of the units at the Browns Ferry site. The licensee has notified the NRC Senior Resident Inspector. See EN #53248. Notified DHS SWO, FEMA Ops, DHS NICC, FEMA NWC (email) and NuclearSSA (email).
The licensee provided additional information regarding the event. Notified the R2DO (Musser), IRD MOC (Gott), NRR EO (Miller). |
| ENS 52025 | 20 June 2016 15:27:00 | This notification is being made pursuant to 10 CFR 50.72(b)(2)(xi) because the Tennessee Valley Authority (TVA) is in the process of informing the Alabama Radiological Protection Department, Alabama Department of Environmental Management, Limestone County Emergency Management Department, and Nuclear Energy Institute (NEI) of recent groundwater monitoring results at the Browns Ferry Nuclear Plant in accordance with NEI 07-07, Industry Ground Water Protection Initiative. There are no indications of any impacts to any off-site drinking water source as indicated by Browns Ferry's off-site groundwater monitoring well samples. TVA has taken immediate action to address the apparent leak following the detection of elevated tritium levels from on-site groundwater monitoring wells and will be monitoring affected wells on an increased frequency. No elevated tritium levels have been detected from off-site monitoring locations, and the public is not at risk. The licensee has notified the NRC Resident Inspector. |
| ENS 48829 | 19 March 2013 08:37:00 | At 0402 (CDT) on 03/19/2013, the Unit 1 reactor was manually scrammed due to lowering main condenser vacuum. The cause of the loss of vacuum was a significant leak on the 1C feedwater heater level control line. The leak appeared as a steam/water leak near the penetration to the main condenser. As extraction steam was isolated, condenser vacuum deteriorated and was approaching the turbine trip setpoint, at which time the reactor was manually scrammed. Condenser vacuum recovered following the scram. MSIVs (Main Steam Isolation Valves) are open, main turbine bypass valves are controlling reactor pressure and reactor feedwater pumps are being used to control reactor water level. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated as required. In response to the scram, all plant equipment responded as designed. The reactor had been operating near 95% power for several hours due to the 1C3 heater isolating at 2334 (hrs. CDT) on 3/18/13. 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.' It is also reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR50.73(a)(2)(iv)(A). All rods inserted into the core during the scram. No safety relief valves lifted during the transient. The plant is in its normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector. |
| ENS 47972 | 29 May 2012 07:22:00 | On 5/29/2012 at 0331 (CDT) the Unit 3 reactor scrammed due to turbine control valve fast closure initiated by a load reject signal on the Main Generator. The cause of the load reject signal is Main Transformer differential relay 387T. Reactor power at the time of the SCRAM was approximately 75%. All systems responded as expected to the load reject signal. Main Steam Isolation Valves remained open and reactor pressure is being controlled on the Main Turbine Bypass Valves. No Main Steam Relief Valves lifted during the transient. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation setpoints were reached. Primary Containment Isolation Signals (PCIS) Groups 2, 3, 6 and 8 were received. The lowest reactor water level observed was -41 inches. Reactor water level was restored to and is being controlled by the Feedwater system in the normal band. 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.' It is also reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR50.73(a)(2)(iv)(A). This event is documented in the station corrective action program on SR# 557947. The NRC Resident Inspector has been notified. All control rods inserted into the reactor core. Electrical power is being back fed from offsite power through the 161 KV feeder line. The reactor is being cooled down within the Technical Specification rates. |
| ENS 47784 | 28 March 2012 22:44:00 | On March 14, 2012, it was determined that in the event of an Appendix R fire, fire damage to cables in Fire Area 2-4, 3-3, or 25-1 could cause a Residual Heat Removal Service Water System (RHRSW) pump to spuriously start. This unanalyzed condition was originally determined to be not reportable. However, during subsequent review on March 28, 2012, this condition was determined to be reportable since the additional load from the spurious start of the RHRSW pump on the associated Emergency Diesel Generator (EDG) could result in the design maximum loading of the EDG being exceeded. As a result of the EDG maximum loading being exceeded, the EDG credited in the Appendix R fire analysis to supply required Appendix R safe shutdown loads could fail. The issue has significant safety impact since the capability to supply power to the Appendix R safe shutdown equipment for the credited EDG is necessary to provide adequate core cooling and decay heat removal during performance of the Browns Ferry Nuclear Plant Safe Shutdown Instructions in an Appendix R fire event. The NRC Resident Inspector has been notified. |
| ENS 47323 | 5 October 2011 19:17:00 | At 0930 on 10/5/11, a fuse failure in a 1E to non - 1E interface failed causing a false run signal to the water treatment system for a standby station service water pump. The water treatment system responded by opening the injection valve for the out of service pump. With the associated pump out of service, the sodium hypochlorite can communicate with the Delaware river. This was discovered at 1553 and immediately terminated. This resulted in the discharge of approximately 195 gallons of 15% sodium hypochlorite to SSW intake which communicates with the Delaware river. The associated service water pump was started thus withdrawing residual product from the intake. A portion of the discharged sodium hypochlorite may have entered the Delaware river. A sample was obtained from the Delaware river just outside the intake structure and there was no detectable chlorine in the sample and there were no abnormal conditions noted on the river. Based on Environmental Licensing review, this was reportable to the state of NJ within 15 minutes. The report to the state was initiated at 1608 on 10/5/11. Subsequent calls were made to the Hope Creek Senior NRC Resident, the National Response Center and the US Coast Guard. Affected systems are limited to the Station Service Water (SSW) system and the sodium hypochlorite water treatment system. There was no impact on plant operations and these systems remain fully operable. Evaluation is underway to fully understand the impacts of the fuse failure to prevent reoccurrence. There was nothing unusual or not understood. All safety related equipment continues to function as required. There were no injuries or reported wildlife impact. |
| ENS 46870 | 21 May 2011 01:56:00 | On 5/20/2011 at 2217 (CDT), while performing 1-SR-3.5.1.1(HPCI) MAINTENANCE OF FILLED HPCI DISCHARGE PIPING, operators opened 1-FCV-73-44, HPCI Injection Valve, to fill and vent portions of the system. Once open, the HPCI discharge piping rapidly pressurized to 1000 psig. Operators immediately shut the 1-FCV-73-44 valve. A flood level alarm was received in the control room and water was confirmed to have been leaking from the Gland Seal Condenser. All leakage has stopped. It is suspected that leakage past 1-CKV-73-45, HPCI TESTABLE CHECK VALVE, caused the rise in discharge piping pressure. An investigation into this event is ongoing. In accordance with TS LCO 3.6.1.3 Condition A, the affected line has been isolated by one closed and deactivated valve. This incident is reportable as an 8-hour ENS notification under 10CFR 50.72 (b)(3)(v) as 'any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: d. Mitigate the consequences of an accident.' It also requires a 60 day written report in accordance with 10CFR 50.73(a)(2)(vii). The NRC Resident Inspector has been notified. The volume of leakage was not specifically known but the leakage resulted in approximately 2 inches of water on the HPCI pump room floor before it was isolated. |
| ENS 45536 | 2 December 2009 17:25:00 | During reconstitution of the post fire safe shutdown analysis, it was noted that conditions exist whereby the requirements of the Hope Creek fire protection program (BTP CMEB 9.5-1) for the required degree of separation for redundant safe shutdown trains may not be met. A postulated fire in one of the reactor building fire zones (RB1 or RB2) may affect the ability to operate a required chilled water circulating pump due to the logic interrelationships between the cooling fans and their associated chilled water pump. Specifically, a fire in one area may affect cooling fans from each division and the fan failures could either trip or prevent the redundant chilled water pump from operating. This has the effect of removing the fire barrier between redundant safe shutdown trains. As such, this event is reportable under 10CFR50.72(b)(3)(ii)(B) for the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. Fire watches have been implemented in accordance with our approved fire protection program as an interim compensatory measure. All required safe shutdown systems remain operable. No systems or safety-related equipment actuations occurred or were required or expected as a result of this condition. All systems and safety-related equipment affected by this condition remain operable and functional. Nothing has been noted as unusual or not understood. The only system affected is the chill water system. There were no actuations or initiating signals, nor effects of the event on the plant. No personnel have been injured. The NRC Resident Inspector will be notified. |
| ENS 44541 | 5 October 2008 12:17:00 | During repair activities associated with the HPCI NUMAC (Nuclear Measurement Analysis And Control) display, the NUMAC drawer failed upon restoration of power. The drawer failure resulted in HPCI isolation signal and (HPCI) turbine trip signal. At the time the HPCI isolation valve was de-energized as planned so the isolation did not occur, but the failure mode resulted in a (HPCI) turbine trip signal, which resulted in an unavailability and inoperability of HPCI. The condition existed for 29 minutes during maintenance activities. Repair activities were unsuccessful and the NUMAC is still inoperable. The event occurred during the restoration of the HPCI steam leak detection system. The licensee will be notifying the NRC Resident Inspector. |
| ENS 41812 | 29 June 2005 17:32:00 | At 1440 hrs (EDT) on 06/29/05, the Emergency Plan Coordinator for PSEG informed the Hope Creek Control Room that 31 offsite sirens were inoperable due to an unplanned power outage. The Hope Creek ECG (Emergency Communications Guide) requires an eight-hour report to be made if greater than 17 offsite sirens are inoperable for more than one hour. All safety systems are available and operable with the exception of the 'C' Safety Auxiliary Cooling Water Pump, and the 'C' Residual Heat Removal Pump, which is cleared and tagged for a scheduled system outage. At 1555 (hrs.), (we were) notified by (the) Emergency Plan coordinator that power was restored to a majority of the sirens in New Jersey. (Nine) sirens in New Jersey are still inoperable, (one) siren in Delaware is still inoperable. The licensee has notified the NRC Resident Inspector. |