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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56531 | 20 May 2023 08:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | 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. | Feedwater Reactor Protection System Main Steam Isolation Valve Primary containment Control Rod | |
| ENS 54795 | 21 July 2020 09:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Biofouling of Intake Structure Leads to Manual Reactor Scram of Unit 1 | 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. | Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Main Condenser | |
| ENS 54794 | 20 July 2020 18:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Biofouling of Intake Structure Leads to Manual Reactor Scram of Unit 2 and Downpower of Units 1 & 3 | 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. | Feedwater Reactor Protection System Main Steam Isolation Valve Primary containment Main Condenser Control Rod | |
| ENS 54302 | 1 October 2019 08:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram During Reactor Startup | 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. | Reactor Protection System Intermediate Range Monitor Emergency Core Cooling System | |
| ENS 53923 | 10 March 2019 04:59:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Scram Resulting in Rps and Eccs Actuation | 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. | Feedwater High Pressure Coolant Injection Reactor Protection System Emergency Diesel Generator Main Steam Isolation Valve Primary Containment Isolation System Reactor Core Isolation Cooling Core Spray Residual Heat Removal Main Steam | |
| ENS 53269 | 18 March 2018 16:58:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Turbine Control Valve Closure | At 1158 CDT on March 18, 2018, the Unit 1 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from High Reactor Steam Dome Pressure in response to Turbine Control Valve Closure. The reactor had been operating at 100 percent power. Investigation is in progress. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Steam Relief Valves (MSRVs) operating on the initial transient as expected. Main Turbine Bypass Valves are currently 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. All safety system operated as expected. At no time was public health and safety at risk. 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 pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours 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) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation' and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. | Feedwater Reactor Protection System Main Steam Isolation Valve Primary containment Control Rod Main Steam | |
| ENS 53162 | 10 January 2018 15:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Turbine Control Valve Fast Closure Scram Signal | At 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. | Feedwater High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Primary containment Control Rod | |
| ENS 52648 | 29 March 2017 23:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram Initiated During Startup | At 1844 CDT on 3/29/2017, Unit 2 initiated a manual scram due to multiple rods inserting. At 1842 during Unit 2 start-up, Intermediate Range Monitor (IRM) 'G' drifted low. The operator adjusted the range down one position with no immediate reaction. At 1844, a spike on IRM 'G' caused a half scram on Reactor Protection System (RPS) 'A' trip system. The half scram was being reset after evaluating no trip condition was present. As the operator reset groups 2 and 3, a trip signal from IRM 'F' was received on the RPS 'B' trip system, resulting in rod insertion for groups 1 and 4. When the operator identified multiple rods inserting, the actions of procedure 2-AOI-100-1 were followed and a manual scram was inserted. Investigation is ongoing. All safety systems remained in standby readiness configuration. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary Containment Isolations Systems did not receive an actuation signal and performed as designed. 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 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 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of systems listed in paragraph (b)(3)(iv)(B) Reactor Protection System(RPS) including reactor scram and reactor trip'. This event requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. | Reactor Protection System Reactor Core Isolation Cooling Primary containment Intermediate Range Monitor Emergency Core Cooling System | |
| ENS 50404 | 26 August 2014 22:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Generator Neutral Overvoltage Causes a Reactor Scram | At 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. | Feedwater High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Reactor Core Isolation Cooling Main Steam | |
| ENS 50090 | 6 May 2014 13:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Low Reactor Water Level During Instrument Testing | At 0830 (CDT) on 05/06/2014, the Unit 3 reactor automatically scrammed due to low reactor water level as a result of a trip of both recirculation pumps. Main Steam Isolation Valves remained open with main turbine bypass valves controlling reactor pressure. Reactor feedwater pumps are in service to control reactor water level. Primary Containment Isolation System Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. The Reactor Feedwater System controlled and maintained water level above the level 2 initiation setpoint. Prior to the Scram, the reactor was operating at 100% power. A Core and Containment Cooling Systems Analog Trip Unit Functional Test was in progress. The cause of the recirculation pump trip is under investigation. 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). The NRC Resident Inspector has been notified. U1 and U2 remained at 100% power and were unaffected.
Investigation revealed that a failed power supply caused an Anticipated Transient Without Scram/Alternate Rod Insertion (ATWS/ARI) signal to be generated when a level 2 Reactor Water Level was simulated on one instrument. All systems responded to the ATWS/ARI signal as designed. This signal opened the Recirc Pump Trip breakers for both Recirculation Pumps and opened the ARI valves to bleed air from the Reactor Protection System (RPS) scram air header. The resulting transient caused reactor water level to dip below the RPS trip setpoint (level 3 Reactor Water Level), a normal plant response, and the automatic scram signal occurred. At the time of the RPS scram signal, all rods were inserting and reactor power was approximately 2-3% and lowering. The NRC Resident Inspector has been notified. Notified R2DO (Bonser). | Feedwater Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System | |
| ENS 49928 | 19 March 2014 03:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Due to Failure to Control Main Turbine Moisture Separator Level | At 2252 on 03/18/2014, the Unit 3 reactor automatically scrammed due to a turbine trip from a high Main Turbine moisture separator level. Initial indications show the level controller for 3B2 Moisture Separator failed to adequately maintain level. Additionally local manual control attempts failed to restore moisture separator level. Main Steam Isolation Valves remained open with main turbine bypass valves controlling reactor pressure. Reactor feedwater pumps are in service to control reactor water level. Primary Containment Isolation System 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. The reactor had been operating near 35% power during scheduled power ascension. 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). NRC Resident Inspector has been notified. | Feedwater High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Reactor Core Isolation Cooling Main Turbine | |
| ENS 48829 | 19 March 2013 09:02:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Lowering Condenser Vacuum | 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. | Feedwater Reactor Protection System Safety Relief Valve Main Condenser | |
| ENS 48782 | 25 February 2013 19:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Automatic Scram Due to a Turbine Trip from a Loss of Condenser Vacuum | At 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. | Feedwater High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Residual Heat Removal Emergency Core Cooling System Safety Relief Valve Control Rod | |
| ENS 48623 | 22 December 2012 17:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Loss of Power to the Reactor Protection System | On 12/22/2012 at 1152 CST, the Unit 2 reactor automatically scrammed due to actuation of the Reactor Protection System (RPS) from loss of power to RPS. At 1134 CST, the D 4kV Shutdown Board unexpectedly de-energized during performance of post-maintenance testing for the 3D Diesel Generator paralleling circuitry, resulting in loss of power to the 2B RPS subsystem. Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations were received along with automatic initiation of A, B, and C Standby Gas Treatment subsystems and A Control Room Emergency Ventilation subsystem due to loss of power to the 2B RPS subsystem. While attempting to reenergize the 2B RPS subsystem, the 2A RPS subsystem was inadvertently de-energized resulting in Unit 2 reactor automatic scram. All affected safety systems responded as expected for the loss of RPS and reactor scram. Due to the loss of RPS, the Main Steam Isolation Valves (MSIVs) closed. Reactor pressure did not rise to the automatic initiation set point for Safety Relief Valve (SRV) actuation. Reactor Core Isolation Cooling System (RCIC) and High Pressure Coolant Injection System (HPCI) reactor water level initiation set point of -45" was reached and RCIC and HPCI automatically initiated as designed to restore water level above the initiation set point. Both Recirculation Pumps also tripped on reactor water level of -45". Reactor pressure control was established by manually operating one SRV and water level control established with RCIC. HPCI was returned to standby readiness. The scram was reset, MSIVs were opened, and the Main Condenser was established as a heat sink. The scram event from critical is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B). 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 was notified. The 2A and 2B RPS subsystems were returned to service. The electrical grid is stable and supplying shutdown loads on Unit 2. Unit 1 and Unit 3 were unaffected and continue to operate at 100% power. | High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Reactor Core Isolation Cooling Control Room Emergency Ventilation Safety Relief Valve Main Condenser | |
| ENS 47972 | 29 May 2012 08:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Main Generator Load Reject Signal | 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. | Feedwater Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Primary containment Main Transformer Emergency Core Cooling System Control Rod Main Steam | |
| ENS 47955 | 24 May 2012 11:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram During Startup | At 0639 CDT on 5/24/2012. Unit 3 initiated a manual scram due to multiple rods inserting. At 0637 CDT during Unit 3 start-up Intermediate Range Monitor (IRM) 'H' was ranged down instead of up resulting in half scram on Reactor Protection System (RPS) 'B' trip system. The half scram was being reset after IRM 'H' was properly ranged. The operator placed the scram reset switch in Group 2/3 position. As the operator reset groups 2 and 3, a spike on IRM 'A' was received on the RPS 'A' trip system, resulting in rod insertion for groups 1 and 4. When the operator identified multiple rods inserting, the actions of procedure 3-AOI-l00-1 were followed and a manual scram was inserted. Investigation is ongoing. All safety systems remained in standby readiness configuration. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary Containment lsolations Systems did not received actuation signals and performed as designed. 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 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 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of systems listed in paragraph (b)(3)(iv)(B) 'Reactor Protection System (RPS) Including reactor scram and reactor trip.' This event requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. | Reactor Protection System Reactor Core Isolation Cooling Primary containment Intermediate Range Monitor Emergency Core Cooling System | |
| ENS 47942 | 22 May 2012 07:49:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Scram Due to Reactor Protection System Being De-Energized | At 0249 CDT on 5/22/2012, Unit 3 reactor automatically scrammed due to de-energization of Reactor Protection System (RPS) from actuation of 3A Unit Station Service Transformer (USST) differential relay 387SA, which resulted in a loss of 500KV power to Unit 3. This relay was picked up during a transfer of 4KV Unit Board 3C from alternate power (161KV) to normal power (3A USST). Investigation is in progress as to the cause of relay actuation. 500KV power was restored through the alternate feeder breakers from 161KV to all Unit 3 4KV Unit Boards successfully. 161KV remained available during the entire event. Loss of 500KV power lasted less than 30 seconds and power has been restored to all safety related boards. All Unit 3 diesel generators successfully started and tied to their respective 4KV Shutdown Boards. All safety systems responded as expected to the loss of 500KV power. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. RCIC was manually started to control reactor water level. Primary Containment Isolation System (PCIS) and PCIS initiation signals for groups 1, 2, 3, 6 & 8 were received as designed. At the time of the scram, High Pressure Coolant Injection (HPCI) system was tagged out for removal of temporary instrumentation following planned maintenance. 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). The NRC Resident Inspector has been notified. | High Pressure Coolant Injection Reactor Protection System Primary Containment Isolation System Reactor Core Isolation Cooling Emergency Core Cooling System | |
| ENS 47299 | 28 September 2011 09:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to a Turbine Trip | At 0414 (CDT) on 9/28/2011, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System (RPS) from a turbine trip. Preliminary indications show the turbine tripped on a generator trip with generator neutral overvoltage (359GN) relay actuation. Cause of relay actuation is under investigation. Seven Safely Relief Valves (SRVs) cycled due to the reactor pressure transient with reactor pressure automatically controlled by the Main Turbine Bypass Valves. 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. Primary containment isolation and initiation signals for groups 2, 3, 6 & 8 were received as expected. Reactor water level is being automatically controlled by the feedwater system. 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 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 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. All control rods fully inserted. The plant is being supplied from offsite power and is in a normal shutdown configuration. The MSIVs are open with decay heat being removed via steam to the main condenser using the bypass valves. There was no impact on Units 1 or 2. | Feedwater Reactor Protection System Reactor Core Isolation Cooling Primary containment Emergency Core Cooling System Main Condenser Control Rod | |
| ENS 46793 | 27 April 2011 22:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Notification of Unusual Event Due to Loss of Offsite Power | At 1701 CDT, the licensee declared a Notification of Unusual Event under Emergency Action Level 5.1U due to loss of offsite power for >15 minutes. The loss of offsite power occurred at 1635 CDT and was due to severe weather and winds in the vicinity. When offsite power was lost, all 3 units automatically scrammed. The units are currently stable in Mode 3 with their respective 4KV busses being supplied by the onsite Emergency Diesel Generators(EDG). The 161KV Athens line is the only offsite power source energized. All onsite safe shutdown equipment is available with the exception of the Unit 3 "B" EDG which was out of service for planned maintenance.
The system actuations that occurred during the loss of offsite power were actuations of the Reactor Protection System, Primary Containment Isolation System (PCIS) and Emergency Diesel Generators. All primary containment valves actuated by the PCIS operated as expected. Unexpectedly, the Unit 3 "B" Main Steam Isolation Valve indicates intermediate. Unit 1 High Pressure Coolant Injection actuated when reactor water level reached -45". Reactor Core Isolation Cooling (RCIC) was already initiated at the time.
Following the loss of offsite power only 12 of the required 100 offsite emergency sirens are operable. The licensee will inform both state/local agencies and the NRC Resident Inspector. Notified R2IRC (Wert) of this update.
As a result of the loss of offsite power, the Diesel-driven Fire Pump auto-started. While the pump was running, the licensee discovered that approximately one quart of oil had leaked from the fire pump into the cold water channel which discharges into navigable waterways. The licensee confirmed this at 1950 CDT by visually identifying a sheen in the channel. The licensee notified the National Response Center of the spill and, in accordance with their site discharge permit, notified the State of Alabama. This constitutes an Offsite Notification in accordance with 10CFR50.72(b)(2)(xi). The licensee notified the NRC Resident Inspector. Notified NRC R2IRC (Wert).
At 2120 CDT, operators on Unit 1 were controlling reactor water level between 2 and 51 inches when RCIC became sluggish and water level dropped to +2" causing a valid RPS Scram signal as well as PCIS signals 2, 3, 6, and 8. All valves operated as expected and all isolations were completed. The licensee notified the NRC Resident Inspector. Notified NRC R2IRC (Wert). * * * UPDATE FROM WILLIAM BAKER TO CHARLES TEAL AT 2338 EDT ON 4/28/11 * * * At 1635 CDT following offsite power grid oscillations (due to inclement weather), and a subsequent Unit 1 power reduction from 100% to 75% to attempt to correct the condition, BFNP experienced a complete loss of the 500kV offsite power system. This resulted in an automatic turbine trip and reactor scram of Units 1, 2 and 3. One 161 kV offsite power system (Athens) remains available. This condition is reportable IAW 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 (4-hour notification). This notification was reported to NRC (Crouch) at 1723 CDT. At 1701 CDT, a NOUE was declared (EAL Designator 5.1-U) due to loss of normal and alternate voltage to all 4kV SD (Shutdown) Boards for greater than 15 minutes and at least two Diesel Generators supplying power to unit specific 4kV SD Boards. This condition is reportable IAW 10CFR 50.72(a)(1)(i) - The declaration of any of the emergency classes specified in the licensee's emergency plan (1-hour notification). This notification was reported to NRC (Crouch) at 1723 CDT. Following the initial scrams, there were valid actuation signals for RPS (U1/2/3), Containment Isolation Groups 2, 3, 6 and 8 (U1/2/3), HPCI (U1 only), and Emergency Diesel Generators A, B, C, D, 3A, 3C and 3D (EDG 3B is out of service for maintenance). MSIV's (U1 and 3) closed on loss of A and B RPS power, and the U3 B inboard MSIV is indicating 'double lit' (not fully closed) at this time. All other systems responded as expected. This condition is reportable IAW 10CFR50.72(b)(3)(iv)(A) - Any event or condition that results in the valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation (8-hour notification). The systems with a valid actuation were RPS, Containment Isolation, HPCI and Emergency Diesel Generators. This was reported in an EN# 46793 update to NRC (Crouch) at 1842 CDT. At 1820 CDT a determination was made that offsite emergency notification sirens did not meet the minimum required number operable. Seventy of the one hundred sirens are required to be operable and twelve of the sirens are operable at this time. This condition is reportable IAW 10CFR50.72(b)(3)(xiii) - Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, emergency notification system, or offsite notification system) (8-hour notification). This was reported in an EN# 46793 update to NRC (Crouch) at 2053 CDT. Following auto-start of the diesel driven fire pump, subsequent to the loss of offsite 500kV power system, approximately one quart of oil leaked from the drain plug in the diesel engine of the pump to the plant cold water channel (waters of the United States). This oil produced a "sheen" on the water (confirmed at 1950 CDT) that required a response to the National Response Center IAW 40CFR112.7(a)(4). This condition was reported to the National Response Center at 2055 CDT and assigned spill number 974232. In addition, IAW the BFNP NPDES (National Pollution Discharge Elimination System) permit, the State of Alabama was notified at 2100 CDT of the spill and subsequent notification of the National Response Center. The notification of these outside agencies is reportable IAW 10CFR50.72(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 of other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials (4 hour notification). This was reported in an EN# 46793 update to NRC (Crouch) at 2203 CDT. At 2120 CDT, Unit 1 received a low reactor water level scram due to reactor water level lowering to +2 inches following sluggish RCIC response at low reactor pressure. At the time of this event RCIC and CRD were injecting to the vessel and the reactor level band specified was +2 to +51 inches. A valid Containment Isolation signal was received and groups 2, 3, 6 and 8 isolated as expected. Water level was immediately restored to within the specified band (RCIC). This condition is reportable IAW10CFR 50.72(b)(3)(iv)(A) - Any event or condition that results in the valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation (8-hour notification). The systems with valid actuations were RPS and Containment Isolation. The Emergency Diesel Generators were already running at the time of the event. This was reported in an EN# 46793 update to NRC (Crouch) at 2238 CDT. The NRC resident has been notified of these 1, 4 and 8 hour reports and EN#46793 updates. These conditions and notifications will be captured in the licensee's Corrective Action Program.
At 2050 CDT, on 05/02/2011, the previously declared and reported NOUE (EAL Designator 5.1U) due to loss of normal and alternate voltage to all 4kV SD Boards for greater than 15 minutes and at least two Diesel Generators supplying power to unit specific 4kV SD Boards was terminated due to the conditions requiring entry being resolved. At this time, offsite power has been restored from two 161kV sources (Athens and Trinity), all eight 4kV SD boards are being powered from offsite sources, and six of eight Emergency Diesel Generators (B, C, D, 3A, 3C, 3D) are operable and in standby readiness. Emergency Diesel Generators A and 3B are not operable but are available at this time. All three units remain shutdown, in Mode 4, pending return of the 500 kV grid system. A timeline for return of the 500 kV grid system is yet to be finalized. In addition, the previous 8-hour notification of offsite emergency sirens not meeting the minimum required is being updated to reflect current conditions. As of 1015 CDT, on 05/02/2011, repair activities have resulted in 92 of 100 sirens being in operable status, thereby meeting the minimum requirement of 70 operable. The licensee has notified the NRC Resident Inspector and the State of Alabama. Notified R2DO (Seymour), NRR EO (Nelson), IRD (Grant), DHS (Daly), and FEMA (Dennis). | High Pressure Coolant Injection Reactor Protection System Emergency Diesel Generator Main Steam Isolation Valve Primary Containment Isolation System Reactor Core Isolation Cooling Primary containment | 05000259/LER-2011-001 |
| ENS 46511 | 26 December 2010 22:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to High Vibration on Main Generator Exciter | On 12/26/2010 at 1620 CST, Browns Ferry Unit 3 initiated a manual reactor SCRAM due to high vibration on the Unit 3 Generator Exciter inboard and outboard journal bearings. All plant systems responded as required to the manual SCRAM signal. No Emergency Core Cooling System (ECCS) initiations occurred as a result of the manual SCRAM signal. Reactor water level lowered below Level 3 (+2") as a result of the SCRAM and was recovered to the normal water level band by the Reactor Feed Pumps (RFPs). The expected Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations were received due to reactor water level lowering below Level 3 (+2") with all systems isolating as required. There were no indications of main steam relief valves (MSRVs) opening. The manual scram from critical is reportable within four hours under 10CFR50.72(b)(2)(iv)(B), 'Any event or condition that results in a valid actuation of the Reactor Protection System' and within eight hours under 10CFR50.72(b)(3)(iv)(A), 'Any event that results in an actuation of the specified systems.' The manual scram from critical also requires a 60-day written report in accordance with 10CFR50.73(a)(2)(iv)(A). The event was entered into the licensee corrective action program as Problem Evaluation Report 301505. The NRC Resident Inspector was informed. All control rods fully inserted. Plant is in a normal post-scram electrical alignment. Decay heat is being removed through the turbine bypass valves to the main condenser. | Reactor Protection System Primary Containment Isolation System Emergency Core Cooling System Main Condenser Control Rod Main Steam | |
| ENS 45990 | 9 June 2010 08:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Due to Closure of Main Steam Isolation Valves | At 0331 CDT on 6/9/10, the Unit 2 reactor automatically scrammed due to closure of the Main Steam Isolation Valves (MSIVs). Operating Instruction 2-OI-99 section 8.1, Reactor Protection System (RPS) Bus B Transfer from Motor Generator to Alternate, was in progress for planned maintenance. The MSIVs closed during the RPS power transfer. The cause of the closure of the MSIVs is under investigation. All systems responded as expected to the reactor scram. Safety Relief Valves (SRVs) opened automatically as designed to limit the pressure transient. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling system (RCIC) reactor water level initiation set points were reached and all expected containment isolation and initiation signals were received. Reactor pressure control was established by manually operating one SRV then maintained using the Main Steam Line Drain Valves. RCIC and the High Pressure Coolant Injection system (HPCI) were manually initiated to control reactor water level. The scram was reset, MSIVs were opened, and the Main Condenser was established as a heat sink. Reactor water level control was established with the Reactor Feedwater System and RCIC and HPCI were returned to standby readiness. At 0408 CDT on 6/9/10, a full scram signal was received when 2F Intermediate Range Monitor (IRM) spiked momentarily followed by a spike on 2C IRM. The reactor was stable and operating in Mode 3, Hot Shutdown. No ECCS or RCIC initiation set points were reached. No additional containment isolation signals or initiation set points were received. The cause of the 2C and 2F IRM spikes is under investigation. The scram event from critical 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). The scram received at 0408 CST is reportable within 8 hours 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), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation,' and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector was notified. All rods fully inserted as a result of the first reactor scram. The plant is currently in a normal, post-trip electrical line-up. All SRVs did reseat. There was no impact to the other two units. * * * UPDATE FROM BILL BAKER TO PETE SNYDER ON 6/10/10 AT 1749 EDT * * * Additional review of available data and inspection results revealed that Safety Relief Valves (SRVs) did not lift automatically during the scram. The only operations of SRVs were performed manually to control reactor pressure until the Main Steam Isolation Valves (MSIVs) were reopened. All other details described in the original event notification remain as stated. The licensee notified the NRC Resident Inspector. Notified R2DO (Nease). | Feedwater High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Intermediate Range Monitor Emergency Core Cooling System Main Steam Line Safety Relief Valve Main Condenser | |
| ENS 45391 | 30 September 2009 04:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Unit 2 Manually Scrammed After Trip of an Operating Condensate Pump and Rapidly Lowering Rvwl | On 9/29/09, at 2323 (hours) Unit 2 was manually scrammed due to loss of one of the remaining two 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. The operating crew was removing feedwater pump 2B from service when the condensate booster pump tripped. The condensate booster pump 2C was already out of service to support maintenance. 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. 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 10CFR50.72(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). Lowest observed Reactor Vessel Water Level (RVWL) was -50 inches. Following actuation of HPCI level recovered to +51 inches and then returned to the normal operating band of +33 inches. Safety-related equipment out-of-service prior to the scram included Core Spray Loop 1. All control rods fully inserted. Unit 2 is in a normal post scram electrical lineup. The licensee informed the NRC Resident Inspector and does not plan a press release.
The initial notification made at 0409 hours ET on September 30, 2009, reported that the RCIC system actuated as expected in conjunction with the HPCI to restore Reactor Pressure Vessel (RPV) water level. However, during a review of plant data, BFN (Browns Ferry Nuclear) determined that after receiving a valid actuation signal, RCIC failed to inject to the RPV. The cause of the failure is under investigation. The licensee informed the NRC Resident Inspector of the update and does not plan a press release. Notified R2DO(Ernstes). | Feedwater Secondary containment Reactor Protection System Reactor Recirculation Pump Reactor Pressure Vessel Core Spray Control Rod Main Steam | |
| ENS 45290 | 24 August 2009 23:50:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Condensate Booster Pump Trip Resulting in a Manual Reactor Scram | 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. | Feedwater Secondary containment Reactor Protection System Reactor Recirculation Pump Main Steam | |
| ENS 44860 | 18 February 2009 09:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Trip Due to Power Load Unbalance Signal on Main Generator Resulting in Reactor Scram | 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) | Feedwater Reactor Protection System Reactor Core Isolation Cooling Control Rod Main Steam | |
| ENS 44854 | 16 February 2009 11:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Loss of Stator Cooling Water | 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. | Reactor Protection System Stator Cooling Water Decay Heat Removal Main Condenser | |
| ENS 44540 | 5 October 2008 03:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Following Reverse Power Signal on Main Generator | On 10/04/08 at 2008 (CDT) the Unit 2 reactor scrammed due to turbine generator reverse power signal on the Main Generator. The cause of the reverse power signal is unknown and the investigation is continuing. All systems responded as expected to the generator reverse power signal. 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 set points were reached, and reactor water level is being automatically controlled by the feedwater system. This report is being made as required by 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The licensee characterized the scram as uncomplicated. All control rods fully inserted. No safety valves lifted during the transient. All safety systems were available at the time of the scram. There were no impacts on Units 1 or 3. The licensee has notified the NRC Resident Inspector. | Feedwater Reactor Core Isolation Cooling Emergency Core Cooling System Control Rod | |
| ENS 43878 | 1 January 2008 03:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 3 Automatic Scram Due to Main Generator Load Reject Signal | 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. | Feedwater Reactor Protection System Reactor Core Isolation Cooling Emergency Core Cooling System Control Rod Main Steam | |
| ENS 43718 | 12 October 2007 13:02:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Turbine Trip | At 0802 (CDT) on 10/12/2007 with Unit 1 at 100% power, an automatic reactor scram was received due to a turbine trip. Unit 2 and 3 were also at 100% power and were unaffected by the event. All expected PClS isolations, Group 2 (RHR Shutdown Cooling), Group 3 (RWCU), Group 6 (Ventilation), and Group 8 (TIP) were received along with the auto start of CREVs and 3 SBGT trains. This event is reportable as a 4-hour and 8-hour non-emergency notification along with a 60-day written report in accordance with 10 CFR50.72(b)(2)(iv)(B), 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.73(a)(2)(iv)(A) as 'any event or condition that results in valid actuation of RPS or PCIS'. All control rods fully inserted, the electrical grid is stable, the EDGs and ESF systems remain available, and decay heat is being removed via the Turbine Bypass Valves. The licensee notified the NRC Resident Inspector. The cause of the turbine trip is under investigation. The 1D RHR pump is OOS for planned maintenance. Isolations occurred as a result of low reactor water level +2 inches. The licensee notified the NRC Resident Inspector. | Shutdown Cooling Control Rod | |
| ENS 43613 | 3 September 2007 07:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Electro-Hydraulic Control (Ehc) System Leak | At 0214 (CDT) on 09/03/2007 with Unit 1 at 100% power a core flow runback and manual reactor scram were initiated due to a Electro-Hydraulic Control (EHC) System leak. Units 2 and 3 were also at 100% power and were unaffected by the event. As expected reactor water level momentarily lowered below +2 inches (Reactor Low Water Level) and all appropriate PCIS (Primary Containment Isolation System) isolations, Group 2 (RHR Shutdown Cooling). Group 3 (RWCU), Group 6 (Ventilation), and Group 8 (TIP) were received along with the auto start of CREVs (Control Room Emergency Ventilation) and 3 SBGT trains. This event is reportable as a 4-hour and 8-hour non-emergency notification along with a 60-day written report in accordance with 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50 72(b)(3)(iv)(A) and 10 CFR 50.73(a)(2)(iv)(A) as 'any event or condition that results in valid actuation of RPS or PCIS.' All control rods fully inserted, the electrical grid is stable, the EDGs and ESF systems remain available, and decay heat is being removed via the Turbine Bypass Valves and reactor water level is being controlled by the Reactor Feedwater System. The licensee notified the NRC Resident Inspector. Investigation into the cause of the leak is ongoing.
The Unit 2 Refuel Zone Inboard Exhaust Damper, 2-FCO-64-10, failed to indicate closed in the Control Room at Panel 2-9-25 in response to the Group 6 isolation signal received during the Unit 1 Reactor SCRAM on 09/03/2007 at 0214 CDT as required by the plant design. The damper was physically verified to be open. The Unit 2 Refuel Zone Outboard Exhaust Damper, 2-FCO-64-09 automatically isolated as designed in response to the Group 6 isolation signal fulfilling its Safety Function, was verified closed and was later tagged closed as required by Technical Specifications. Investigation into the cause of the failure of the 2-FCO-64-10 is ongoing. The licensee notified the NRC Resident Inspector. R2DO ( Ernstes) notified. | Feedwater Shutdown Cooling Control Rod | |
| ENS 43560 | 11 August 2007 22:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Scram Due to Neutron Monitoring Trip Signal | On 08/11/2007 at 1751 CDT, Browns Ferry Unit 1 received an Automatic SCRAM due to a Neutron Monitoring (APRM) Trip Signal . Preliminary investigation indicates the trip signal was caused by a Recirculation System Flow Transmitter sensing line becoming separated giving an indicated low flow signal to the neutron monitoring system. With the indicated low flow and high (100%) power, the neutron monitoring system initiated an APRM Simulated Thermal Power Flow Biased Reactor Scram. All control rods inserted and all systems responded as required to the automatic SCRAM signal. No Emergency Core Cooling System (ECCS) initiations occurred as a result of the SCRAM. Reactor water level lowered below Level 3 (+2") (lowest indicated level reached -33") as a result of the SCRAM and was recovered to the normal level band by the Reactor Feed Pumps (RFPs). The expected Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolations were received due to Reactor Water Level lowering below Level 3 (+2") (lowest indicated level reached -33") with all systems isolating as required. Reactor pressure is being controlled using Main Steam Bypass Valves. Reactor Level is being maintained in band using Reactor Feed Pumps. Plant to remain in Mode 3 and initiate repairs to the failed sensing line. Investigation into the event is proceeding. This event is reportable under 10CFR50.72(b)(2)(iv)(B), any event or condition that results in a valid actuation of the Reactor Protection System; 10CFR50.72(b)(3)(iv)(A), any event that results in an actuation of the specified systems. This event also requires a 60 day written report in accordance with 10CFR50.73(a)(2)(iv)(A). The NRC Senior Resident Inspector has been informed of this event. The sensing line had a flow limiter on it and the line was isolated locally. Amount of leakage not known at this time. | Reactor Protection System Primary Containment Isolation System Emergency Core Cooling System Control Rod Main Steam | |
| ENS 43414 | 9 June 2007 16:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due Turbine Trip as a Result of High Moisture Separator Reheater Tank Level | On 06/09/2007 at 1100 CDT, Browns Ferry Unit 1 received an automatic SCRAM due to a Turbine Trip Signal caused by a Moisture Separator Drain Tank Level High. All control rods inserted and all systems responded as required to the automatic SCRAM signal. Two Main Steam Relief Valves (MSRVs) momentarily lifted in response to the pressure transient. No Emergency Core Cooling System (ECCS) initiations occurred as a result of the SCRAM. Reactor water level lowered below Level 3 (+2") as a result of the SCRAM and was recovered to the normal level band by the reactor feed pumps (RFPs). The expected Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolations were received due to reactor water level lowering below Level 3 (+2") with all systems isolating as required. Additionally, the expected initiation of the RPT breakers from the turbine trip was received which resulted in the trip of both reactor recirculation pumps. Reactor pressure is being controlled using Main Steam Bypass Valves. Reactor Level is being maintained in band using RFPs. Cooldown is in progress to Mode 4. Investigation into the event is proceeding. The NRC Senior Resident Inspector has been informed of this event. This event is reportable under 10CFR50.72(b)(2)(iv)(B) 'any event or condition that results in a valid actuation of the Reactor Protection System'; 10 CFR50.72(b)(3)(iv)(A), ' Any event that results in an actuation of the specified systems'. This event also requires a 60 day written report in accordance with 10CFR50.73(a)(2)(iv)(A).
Follow-up review of the reported reactor scram revealed that the Group 8 isolation did not function as required. Specifically, the licensee provided the update below concerning the function of one of the reactor's five Traversing Incore Probes (TIP) that had been used the previous day for flux mapping and were in the drywell (not fully retracted) to permit decay when the scram occurred. The Group 8 isolation signal received during Unit 1 Rx SCRAM on 06/09/2007 @ 1100 did not automatically go to completion as designed. The 'D' TIP failed to automatically withdraw as required. When the TIP was manually withdrawn, the TIP Ball valve closed as required. The local resident was notified. A work order and PER was written to correct the deficiency. The other four TIPs did retract and the corresponding ball valves shut as expected. R2DO (Fredrickson) notified.
Review of available data indicates that no Main Steam safety relief valves (MSRVs) opened in response to the Unit 1 reactor scram on 06-09-2007. There were no indications of an open MSRV on any discharge tailpipe thermocouple or acoustic monitor. Initial indications of the discharge tailpipe thermocouples for MSRVs 1-PCV-1-5, 1-PCV-1-30, and 1-PCV-1-31 did indicate slight increases in temperature (5 to 18 degrees F) as reactor pressure decreased, which resulted in the initial assumption of two SRVs opening. However, this behavior is a classical indication of slight main seat leakage. This equipment condition is under review via an open PER. The multiple reactor pressure instrumentation responses were reviewed. The peak reactor pressure was indicated at approximately 1093 psig which is 42 psi below the lowest nominal MSRV setpoint. Based upon the observed peak reactor pressure and no indication of an MSRV opening, it would appear that the MSRVs performed as required to during the reactor pressure transient event. The licensee will notify the NRC Resident Inspector. Notified R2DO(Shaeffer). | Reactor Protection System Primary Containment Isolation System Reactor Recirculation Pump Emergency Core Cooling System Safety Relief Valve Control Rod Main Steam | |
| ENS 43381 | 24 May 2007 07:11:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Electro-Hydraulic Control (Ehc) System Leak | On 05/24/2007 at 0211 CDT Browns Ferry Unit 1 initiated a Manual reactor SCRAM due to an Electro-Hydraulic Control (EHC) System pressure lowering and reservoir level lowering due to an EHC system leak. The leak was from #6 Main Turbine Combined Intermediate Valve (CIV). All Systems responded as required to the manual SCRAM signal. No Emergency Core Cooling System (ECCS) initiations occurred as a result of the manual SCRAM signal. Reactor water level was maintained in the normal band during the SCRAM. There were no Primary Containment Isolation signals received during the SCRAM. The EHC leak was stopped due to reservoir level depletion and EHC pumps being secured. There were no indications of main steam relief valves (MSRVs) opening. Reactor pressure is being controlled using Main Steam Line Drains. Reactor Level is being maintained in band using Control Rod Drive pumps. Repair of the EHC leak is in progress. The Scram was characterized as uncomplicated. All rods fully inserted. The only significant equipment out of service at the time was RCIC. When the leak was initially discovered, it was about 60 drops per minute. When repairs were attempted, the piping separated and approximately 600 gallons of EHC fluid was discharged out the break onto the turbine building floor. Cleanup of the EHC fluid is in progress and environmental monitoring is in place to assure no offsite release of the spill. The licensee notified the NRC Resident Inspector. | Primary containment Main Turbine Main Steam Line Control Rod Main Steam | |
| ENS 43159 | 9 February 2007 18:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Scram Due to Low Reactor Water Level | At 1208(CST) on 02/09/2007 with Unit 3 at 100% power, an automatic reactor scram (RPS) was received due to lowering water level. Loss of water level was due to lowering condensate flow which in turn caused a reduction in feedwater flow. Reactor water level lowered to -45 Inches. High Pressure Coolant Injection and Reactor Core Isolation Cooling systems initiated as expected. Additionally, the Recirculation Pump breakers tripped as expected. All expected Primary Containment Isolations, Group 2 (RHR Shutdown Cooling), Group 3 (RWCU), Group 6 (Ventilation), and Group 8 (TIP) were received along with the auto start of Control Room Emergency Ventilation (CREV) and 3 Standby Gas Treatment (SBGT) trains. Unit 2 was at 80% power and was unaffected by the event. Investigation has been initiated as to the cause of the lowering condensate flow. This event is reportable as a 4-hour and 8-hour non-emergency notification in accordance with 10 CFR50.72(b)(2)(iv)(B) as any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical, 10 CFR 5172(b)(3)(Iv)(A) as any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(Iv)(b), and 10 CFR 50.72(b)(2)(Iv)(A) as any event or condition that results or should have resulted In ECCS discharging to the reactor coolant system. All control rods fully inserted, the electrical grid is stable. Decay heat is being removed via the turbine bypass valves to the main condenser. The licensee has notified the NRC Senior Resident Inspector. There was no excessive cooldown rate during the injection phase as cooldown rate did not exceed 100 degrees Fahrenheit. Total injection time for both HPCI and RCIC systems was approximately 2 minutes which resulted in approximately 13,000 gallons of coolant from the condensate storage tank (CST) entering the reactor vessel. Primary plant temperature and pressure are 531 degrees Fahrenheit and 928 psig, respectively. | Reactor Coolant System Feedwater High Pressure Coolant Injection Reactor Protection System Reactor Core Isolation Cooling Primary containment Shutdown Cooling Control Room Emergency Ventilation Main Condenser Control Rod | |
| ENS 43092 | 11 January 2007 14:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram When Turbine Output Breakers Opened | Unit 2 reactor scrammed due to a main turbine trip. PCIS (Primary Containment Isolation Signal) groups 2, 3,and 6 isolated, CREV (Control Room Emergency Ventilation) A, SBGT (Standby Gas Treatment) trains A, B, and C started as expected. All control rods fully inserted, eight main steam relief valves lifted and no ECCS actuations occurred. This event is reportable within four hours according to 10CFR50.72(b)(2)(iv)(B) and eight hours according to 10CFR50.72(b)(3)(iv)(A). The turbine trip was the result of the 500kV main output breaker opening causing the generator output breaker to open. All relief valves fully seated during the scram recovery. The reactor water level is being maintained using normal feedwater and decay heat is being removed using the steam dumps. The plant is using the startup transformer for electrical power. The cause of the 500kV breaker opening is under investigation. The licensee will notify the NRC Resident Inspector. | Feedwater Control Rod Main Steam | |
| ENS 42813 | 30 August 2006 03:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Electro-Hydraulic Oil Leak | On 8/29/06 at 2225 CDT, Browns Ferry Unit 3 initiated a Manual reactor SCRAM due to EHC (Electro-Hydraulic Control) System Reservoir level lowering due to a EHC system leak. The leak was from #2 Main Turbine Control Valve. All Systems responded as required to the manual SCRAM signal. No ECCS (Emergency Core Cooling Systems) initiations occurred as a result of the manual SCRAM signal. Groups 2 (floor drains, etc.), 3 (Reactor Water Cleanup), 6 (Ventilation), & 8 (TIPs) PCIS isolations occurred at + 2 (inches) as expected as a result of the manual SCRAM with all systems isolating as required. The EHC leak rate lowered to approximately zero upon turbine trip. No indications existed of main steam relief valves (MSRVs) opening. Bypass valves controlled reactor pressure due to EHC system staying in service. Repair of the EHC leak is in progress. This event is reportable under 10CFR50.72(b)(2)(iv)(B), 'any event or condition that results in a valid actuation of the Reactor Protection System'; 10CFR50.72(b)(3)(iv)(A), 'Any event that results in an actuation of the specified systems'. This event also requires a 60 day written report in accordance with 10CFR50.73(a)(2)(iv)(A). Reactor Power was reduced to 78% before the reactor was manually scrammed and all rods fully inserted. The EHC oil is being cleaned up and the oil does not pose a fire threat. All ECCS and the EDGs are fully operable if needed and the electrical grid is stable. The NRC Resident Inspector was notified of this event by the licensee. | Reactor Protection System Reactor Water Cleanup Emergency Core Cooling System Main Steam | |
| ENS 42787 | 19 August 2006 16:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram of Unit 3 Due to the Loss of Both Reactor Recirculation Pumps | At 1105 CST both Reactor Recirculation pumps tripped on Unit 3. In accordance with procedure 3-AOI-68-1, a manual scram was initiated. Reactor water level lowered to Reactor Pressure Vessel Level 3, resulting in the automatic actuation of the Primary Containment Isolation System as expected: Group 2 (RHR Shutdown Cooling), Group 3 (Reactor Water (Clean Up), Group 6 (Ventilation), and Group 8 (Traversing Incore Probe) along with the automatic start of Control Room Emergency Ventilation and all 3 trains of the Standby Gas Treatment System. Reactor water level was recovered to normal levels with the reactor feedwater system. During this time Unit 2 was at 100% power and was unaffected by the event. This event is reportable as a 4-hour and 8-hour notification along with a 60-day written report in accordance with 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.73(a)(2)(iv)(A). All control rods fully inserted on the scram. Decay heat is being removed with normal feedwater and the turbine bypass valves. No relief valves lifted during this event. Electrical power to the plant is aligned for the normal shutdown lineup. The cause of this event is under investigation. The licensee notified the NRC Resident Inspector. | Feedwater Primary Containment Isolation System Shutdown Cooling Reactor Recirculation Pump Reactor Pressure Vessel Standby Gas Treatment System Control Room Emergency Ventilation Control Rod | |
| ENS 42102 | 31 October 2005 19:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Trip of Main Turbine | At 1318 (CST) on 10/31/05 with Unit 3 at 100% power, a full reactor scram signal (RPS) was received due to a turbine trip. Unit 2 was also at 100% power and was unaffected by the event. Reactor water level lowered to approximately minus 6 inches as expected and was recovered with normal feedwater flow. All expected PCIS isolations, Group 2 (RHR Shutdown Cooling), Group 3 (RWCU), Group 6 (Ventilation), and Group 8 (TIP) were received along with the auto start of CREVs and 3 SBGT trains. This event is reportable as a 4-hour and 8-hour non-emergency notification along with a 60-day written report in accordance with 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.73(a)(2)(iv)(A) as 'any event or condition that results in valid actuation of RPS or PCIS'. The licensee stated that the turbine trip was most likely caused by bus transfer evolutions in the 500 kv switchyard. The licensee stated that all control rods fully inserted, the electrical grid is stable, the EDGs and ESF systems remain available, and that decay heat is being removed via the turbine bypass valves to the main condenser. The licensee notified the NRC Resident Inspector. | Feedwater Main Turbine Shutdown Cooling Main Condenser Control Rod | |
| ENS 41997 | 17 September 2005 16:29:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Due to Maintenance | On 09/17/2005 at 1129 CDT, Browns Ferry Unit 3 reactor automatically scrammed due to the trip of the main generator turbine on low main condenser vacuum. Prior to the trip, the unit was operating at reduced power (74 percent) and maintenance was in progress to repair in place a secondary plant moisture separator level control valve. This maintenance activity resulted in the low condenser vacuum condition. Plant response to the scram was as expected. Post-scram transient RPV low water level caused a valid automatic initiation of primary containment isolation (PCIS) valves group 2 (containment sumps and shutdown cooling valves), group 3 (reactor water cleanup system), group 6 (ventilation systems, including initiation of standby gas treatment and control room emergency ventilation systems), and group 8 (traversing in-core probe system). This event is reportable to the NRC via ENS within four hours under 10CFR50.72(b)(2)(iv)(B) (RPS actuation while critical), and within eight hours under 10CFR50.72(b)(3)(iv)(A) (RPS and containment isolation actuation). Browns Ferry Unit 2 was not affected. The reactor shutdown and all control rods fully inserted. Condenser vacuum has been restored and decay heat is being removed using the bypass valves. | Primary containment Shutdown Cooling Reactor Water Cleanup Control Room Emergency Ventilation Main Condenser Control Rod | |
| ENS 41896 | 5 August 2005 22:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Rps Actuation Due to Low Reactor Water Level Caused by Loss of Reactor Feed Pumps | At 1708 on 08/05/2005, Unit 2 reactor scrammed from 100% power when reactor water level reached Level 3 Reactor Scram setpoint due to loss of 2C and 2B Reactor Feed Pumps. All rods inserted (fully). Unit 3 was also at 100% power and was unaffected by this event. Reactor water level reached level 2 (-45) HPCI & RCIC initiation setpoint and was recovered by HPCI and RCIC injection. All expected PCIS isolations, Group 2 (RHR S/D cooling), Group 3 (RWCU), Group 6 (ventilation), and Group 8 (TIP) were received along with the auto start of CREV, 3 SGT trains, HPCI and RCIC with injection into Reactor Vessel. Recirc Pumps tripped at -45" as expected. Four MSRVs lifted momentarily to stabilize reactor pressure. This event is reportable at 4 hour and 8 hour Non-Emergency Notification along with a 60 day written report in accordance with 10CFR50.72(b)(2)(iv)(A), 10CFR50.72(b)(2)(iv)(B), 10CFR50.72(b)(3)(iv)(A) and 10CFR50.73(a)(2)(iv)(A) as 'Any event or condition that results in a valid actuation of RPS and PCIS as described in (l) and (2) below.' Decay heat is being removed via steam to the main condenser using the bypass valves maintaining pressure at 895 psi. The licensee notified the NRC Resident Inspector.
Upon further evaluation by the trip investigation team, it has been determined that no MSRV's lifted during this event. Reactor pressure data indicated no increasing pressure spike during the transient. The post-scram pressure trends indicated the maximum reactor pressure was normal operating pressure at the start of the transient followed by a gradual decline. The initial report which stated that 4 MSRV's had lifted was in error. The licensee will notify the NRC Resident Inspector. Notified the R2DO (Landis). | Main Condenser | |
| ENS 41404 | 11 February 2005 22:29:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to a Load Rejection | The following text was obtained from the licensee via facsimile: At 1629 (hrs. CST) on 02/11/05, Unit 3 reactor scrammed from 100% power when the output breaker tripped causing a load reject. The breaker tripped due to a corresponding switchyard breaker 5268 tripping when a PK block was re-installed before the trip cutout (TCO) switches were placed in TCO. All rods inserted. Unit 2 was also at 100% power and was unaffected by this event. Water level lowered to +1" as expected and was recovered by normal feed water flow. All expected PCIS (Primary Containment Isolation System) isolations, Group 2 (RHR S/D (Residual Heat Removal) cooling), Group 3 (RWCU (Reactor Water Clean Up)), Group 6 (ventilation), and group 8 (TIP (Transverse Incore Probe)) were received along with the auto start of CREV (Control Room Emergency Ventilation) and the 3 SGT (Standby Gas Treatment) trains. Four MSRV's (Main Steam Safety Relief Valves) lifted momentarily to stabilize reactor pressure. This event is reportable as a 4-hour and 8-hour Non-Emergency Notification along with a 60-day written report in accordance with 10CFR50.72(b)(2)(iv)(B), 10CFR50.72(b)(3)(iv)(A) and 10CFR50.73(a)(2)(iv)(A) as 'Any event or condition that results in a valid actuation of RPS (Reactor Protection System) and PCIS .. The plant was performing restoration from switchyard maintenance at the time of the scram. All safety relief valves that lifted properly re-seated. Decay heat is being removed via the steam bypass valves to condenser. The electrical grid is stable. The licensee has notified the NRC Resident Inspector. | ||
| ENS 41219 | 23 November 2004 16:02:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Turbine Trip | At 1002 on 11/23/2004 with Unit 3 @ 100% power, a full reactor scram signal (RPS) was received due to a Turbine Trip. Unit 2 was also @ 100% power and was unaffected by the event. Reactor water level lowered to approximately -30 (inches) as expected and was recovered with normal feed water flow. All expected PCIS ISOLATIONS, Group 2 (RHR S/D Cooling), Group 3 (RWCU), Group 6 (Ventilation) & Group 8 (TIP) were received along with the auto start of CREVS and the three SBGT Trains. One PCIS Scram Discharge Volume Drain (85-37F) failed to close but the flowpath did isolate. The Unit remains Shutdown and an investigation is underway to determine the cause of the Turbine Trip and resulting Reactor Scram. This event is reportable as a 4-hour & 8-hour Non-Emergency Notification along with a 60 day written Report in accordance with 10CFR50.72(b)(2)(iv)(B), 10CFR50.72(b)(3)(iv)(A) & 10CFR50.73(a)(2)(iv)(A) as 'Any event or condition that results in valid actuation of RPS & PCIS . . . The licensee informed the NRC Resident Inspector. | ||
| ENS 40862 | 11 July 2004 03:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to an Upscale Trip on Intermediate Range Monitors | On 07/10/2004 at 2235 (CST), during Browns Ferry Unit 2 startup activities, as IRMs (Intermediate Range Monitors) were being ranged up, an upscale trip on IRM E (RPS (Reactor Protection System) A Channel) and IRM F (RPS B Channel) was received, resulting in a full reactor scram. Mode Switch was in STARTUP, Mode 2 at time of trip. IRMs were on ranges 6 and 7, and reactor pressure was approximately 950 psig. All systems responded as designed, all control rods are at full-in. No ECCS (Emergency Core Cooling System) or PCIS (Primary Containment Isolation System) actuation set points were reached. This is reportable as 4 hour ENS (Emergency Notification System) report 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 pre-planned sequence during testing or reactor operation.' It is also reportable as an 8 hour ENS 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 and reactor trip.' Also reportable as a sixty day written report per 10CFR50.73(a)(2)(iv)(B). Mode Switch is presently in shutdown, Mode 3. Investigation is still on going. NRC Resident (Inspector) was notified at approximately 2310 (CST). | Reactor Protection System Intermediate Range Monitor Control Rod | |
| ENS 40858 | 9 July 2004 03:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Due to a Turbine Generator Load Rejection Signal | The following information was received from the licensee via facsimile: On 7/08/2004 at approximately 2232 hours CDT, Browns Ferry Unit 2 scrammed due to a Turbine Generator Load Reject Signal. All systems responded as required to the scram signal. No ECCS initiations occurred as a result of this event. Reactor Water Level lowered to the Low Level setpoint which generated a redundant SCRAM signal and initiated the Primary Containment Isolation System (PCIS) function for PCIS groups 2 (Primary Containment), 3 (RWCU) (Reactor Water Clean Up), 6 (Secondary Containment), and 8 (TIP system) (Transverse Incore Probe). As expected, Main Steam Relief Valves (MSRVs) opened due to the high reactor pressure (maximum value observed was 1137 psig) as a result of the Main Turbine / Generator Trip. Reactor Water level was restored to normal via the Reactor Feedwater system and all PCIS isolations have been reset. The cause of the Load Reject Signal is still under investigation at this time. This event is reportable under 10CFR50.72(b)(2)(iv)(B), 'Any event or condition that results in a valid actuation of the Reactor Protection System (RPS) when the reactor is critical' and also under 10CFR50.72(b)(3)(iv)(A), 'Any event or condition that results in a valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B)'. This event also requires a 60 day written report in accordance with 10CFR50.73(a)(2)(iv)(A). All control rods inserted during the scram. All Main Steam Relief Valves have properly reset. Decay heat is being removed from the reactor via main turbine bypass valves to the main condenser. Pressure is being maintained at normal operating pressure. The electrical grid is stable. Unit 3 was not affected by the scram on Unit 2. The licensee has notified the NRC Resident Inspector. | Feedwater Secondary containment Reactor Protection System Primary Containment Isolation System Primary containment Main Turbine Main Condenser Control Rod Main Steam |