Semantic search

Jump to navigation Jump to search
 Entered dateEvent description
ENS 5316210 January 2018 13:53:00At 0928 CST on January 10, 2018, the Unit 3 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from Turbine Control Valve Emergency Trip System pressure low. The reactor had been operating near 73 percent power for an emergent issue for Turbine Control Valve (TCV) No. 3. With TCV No. 3 out of service and closed, the unit was operating with RPS in a half scram condition. A subsequent failure of the TCV No. 2 sensing line resulted in RPS coincidence logic being met for TCV fast closure SCRAM. The investigation of the TCV No. 2 sensing line failure continues. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Turbine Bypass Valves controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. This event is reportable within 4 hours 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 preplanned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident inspector has been notified.
ENS 5040426 August 2014 21:24:00At 1730 CDT on August 26, 2014, Browns Ferry Unit 1 experienced a turbine trip resulting in an automatic reactor scram. The cause of the turbine trip was a control valve fast closure signal that was generated by a turbine trip on generator neutral over voltage signal. The Main Steam Isolation Valves (MSIVs) remained 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 required components actuated as required with the exception of Standby Gas Treatment (SBGT) train A, which is under a clearance for planned maintenance. Neither High Pressure Coolant Injection (HPCI) nor Reactor Core Isolation Cooling (RCIC) initiation signals were received. Initially, three Main Steam Relief Valves (MSRVs) opened to control the pressure surge and subsequently reclosed. 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, 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. Service Request 926468 was initiated in the Corrective Action Program. The plant is in its normal shutdown electrical lineup. The licensee is investigating the cause of the generator neutral overvoltage signal. There was no impact on units 2 and 3.
ENS 5027713 July 2014 21:35:00On 07/13/2014 at approximately 1740 (CDT), it was reported that oil sheen was visible in the intake fore bay. Upon further investigation it was determined that a high cold water channel level had caused diesel fuel oil to wash free from an oil drain device from the channel diesel fire pump into the intake fore bay. This oil spill is reportable to the EPA (National Response Center under 40CFR112). Notification to the National Response Center was made at 1750 (CDT). Notification number is 1088951. A total of <1 gallon was reported to enter US waters. The source of the leak into the intake fore bay and cold water channel has been stopped. All reportable oil spillage has been contained in the intake fore bay (considered US waters). Additional notifications have been made to Alabama Emergency Management. This event requires a 4 hour report pursuant to 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 NRC Resident Inspector has been notified. Notified DOE, EPA, USDA, HHS, and FEMA.
ENS 498662 March 2014 22:10:00On March 2, 2014, during an extent of condition review for a separate problem evaluation report (PER) for the Browns Ferry Nuclear Plant (BFN), it was discovered that a postulated worst case failure hot short of cable PP679-IA associated with 4kV Shutdown Board 'A' cross-tie breaker 1824 may cause spurious opening of breaker 1824. 4KV Shutdown Board 'A' breaker 1824 is required to be closed during an Appendix R safe shutdown event in fire area 2-3. The action taken by 0-SSI-2-3 is to assure breaker 1824 is closed after placing '43 switch' (Breaker Control Transfer switch) and '43AR switch' (Appendix R Isolation switch) in the emergency position. Cable PP679-IA is routed in fire area 2-3 and fire damage resulting in a hot short could prevent closing breaker 1824 upon demand. Therefore, this condition could result in a loss of power to credited safe shutdown equipment used for Unit 1 that would challenge the ability to provide adequate core cooling during performance of BFN Safe Shutdown Instructions. 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). This item has been entered into the Corrective Action Program as PER# 853503. The NRC Resident Inspector has been notified of this event. The licensee has established compensatory actions to ensure breaker 1824 remains closed.
ENS 4963916 December 2013 22:29:00A circuit analysis review for Appendix R Operator Manual Action deficiency extent of condition identified that fire damage to cable ES194-I is not isolated by the local control power transfer switch utilized in the Safe Shut-down Instruction. Fire damage to the non-isolated cable ES194-I in Fire Areas 01-03, 02-01, and 02-03 could cause the RHR Pump 2C to spuriously start (or restart after the Operator local trip action) when 4kV Shutdown Board B is credited for these Fire Areas. An undesired spurious start of RHR Pump 2C could overload the credited Diesel Generator or take away the necessary load capacity to allow operation of other Appendix R fire safe shutdown credited loads. The fire damage postulated would require a short to ES194-I from a separate cable conductor energized with the positive potential of the battery supplying 4kV Shutdown Board B (i.e., normally Shutdown Battery B). It is postulated for a fire-event that the necessary short to ES194-I could come from a cable-to-cable short or from a short to ground as the fire event may cause a separate conductor energized with the positive potential of the associated battery to short to ground. Similar conditions also exist for: RHR Pumps 1A, 1B, 1D, 2A, 2B, 3A, and 3C due to fire damage to cables in one or more Fire Areas. Compensatory actions in the form of an Operator Work Around (OWA) to remove the affected RHR Pump breaker close circuit control power fuses during the affected Safe Shut-down Instructions, a caution order on the appropriate transfer switches referencing the OWA, and fire watches in the affected Fire Areas to mitigate this condition are in place in accordance with the BFNP (Browns Ferry Nuclear Plant) Fire Protection Report. This condition is being reported pursuant to 10CFR50.72(b)(3)(ii)(B) and 10CFR50.72(b)(3)(v). The NRC Resident Inspector has been notified. The licensee is also reporting under 10CFR50.72(b)(3)(v)(D) Accident Mitigation.
ENS 4878225 February 2013 17:49:00At 1313 (CST) on 02/25/2013, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip. Preliminary indications show the turbine tripped on low condenser vacuum. Cause of loss of condenser vacuum has been identified as Reactor Feedwater recirculation piping separation. Main Steam Isolation Valves (MSIVs) were manually closed to isolate the leak. None of the Safety Relief Valves (SRVs) automatically cycled during the transient, and one Safety Relief Valve (SRV) was manually operated to maintain Reactor Pressure due to the Main Turbine Bypass Valves unavailability because of loss of condenser vacuum. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC), reactor water level initiation set points were reached. Reactor water level is being controlled by the RCIC system and Reactor Pressure is being controlled with the High Pressure Coolant Injection (HPCI) system. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated, with the exception of one valve in Group 6. Drywell Continuous Air Monitor (CAM) Inboard Return Isolation Valve 3-FSV-90-257 did not have indication following isolation signal and was not able to be verified locally. Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM was manually isolated at 1422 (CST) with positive indication of isolation, and isolation valves deactivated at time 1514 (CST) to satisfy TS LCO 3.6.1.3 required actions. 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 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). At 1415 (CST), Suppression Pool Water level exceeded -1 inch due to operation with HPCI in pressure control mode, and required entry into TS LCO 3.6.2.2 condition A to restore level within 2 hours. Efforts are being made to lower suppression pool water level within limits. At 1615 (CST), water level remains above -1 inch requiring entry into TS LCO 3.6.2.2 condition B requiring action to be in MODE 3 in 12 hours and MODE 4 within 36 hours. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. All control rods fully inserted and electrical offsite power is in a normal shutdown configuration. Residual Heat Removal is aligned for suppression pool cooling. There was no impact on either Unit 1 or 2.
ENS 4869022 January 2013 12:45:00This 60-day telephone notification is being made in accordance with 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 multiple Main Steam Isolation Valves (MSIVs). On November 23, 2012 at 0435 Central Standard Time, during performance of Surveillance Instruction 1-SI-3.3.1.A, ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping, as the Residual Heat Removal Loop II Shutdown Cooling was being placed in service, Group 1, Division II Primary Containment Isolation System (PCIS) logic groups A2 and B2 actuated resulting in an unanticipated Division II, Group 1 Complete Isolation and subsequent Inboard MSIV closure. The Outboard MSIVs had been previously tagged closed. Plant conditions which initiate Group 1 actuations are Reactor Vessel Low-Low-Low Water Level, Main Steamline Break, and Low Main Steamline Pressure at the Inlet to the Turbine. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The affected equipment responded as designed. This condition was the result of the reactor vessel water level being within two inches of the reactor head vent when Shutdown Cooling was placed into service, causing pressure perturbations. When these perturbations occurred, they gave an indication of low water level, causing the isolation and MSIV closure. 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 (Browns Ferry Nuclear Plant) Corrective Action Program as Problem Evaluation Report 646607. The NRC Resident Inspector has been notified.
ENS 480068 June 2012 11:32:00This 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 April 11 , 2012, at 1515 hours Central Daylight Time (CDT), Operations personnel attempted to transfer the 3B 480V Reactor Motor Operated Valve (RMOV) Board to its alternate power supply for post maintenance testing on the alternate feeder breaker. The 3B 480V RMOV Board failed to transfer to the alternate power supply. Power to the 3B Reactor Protection System (RPS) Bus was lost resulting in a half-scram and actuation of Primary Containment Isolation System (PCIS) Group 6 with the initiation of all three trains (A, B, and C) of Standby Gas Treatment and the initiation of Train 'A' of the Control Room Emergency Ventilation System. Plant Conditions, which initiate PCIS Group 6 actuations, are Low Reactor Vessel Water Level, High Drywell Pressure, High Reactor Building Vent Radiation, or High Refuel Zone Radiation. At the time of the event, these conditions did not exist; therefore, the partial actuations were invalid. The affected equipment responded as designed. On April 11, 2012, at 1520 hours CDT, Operations personnel reset the half-scram from the loss of the 3B RPS Bus. This condition is the result of a bad connection of the breaker to panel contacts due to alignment and/or infrequent manipulation. There were no safety consequences or impact to the health and safety of the public as a result of these events. This event was entered Into the Corrective Action Program as Problem Evaluation Report 535537. The NRC Resident Inspector has been notified of this event.
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BOriginated-20by::Neel-20Shukla-5D-5D-20-5B-5BCategory:ENS-20Notification-5D-5D/-3FEntered-20date/-3FEvent-20description/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3DOpened:/default%3DOpened*:"