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The query [[Category:ENS Notification]] [[Site::Browns Ferry]] [[Reporting criterion::10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation||10 CFR 50.72(b)(3)(iv)(A), System Actuation]] was answered by the SMWSQLStore3 in 0.2475 seconds.


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 Start dateReporting criterionTitleEvent descriptionSystemLER
ENS 543021 October 2019 08:07:0010 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 StartupOn 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
ENS 5392310 March 2019 04:59:0010 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 ActuationAt 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.Primary Containment Isolation System
High Pressure Coolant Injection
Reactor Core Isolation Cooling
Residual Heat Removal
Reactor Protection System
Main Steam Isolation Valve
Core Spray
Feedwater
Emergency Diesel Generator
ENS 5326918 March 2018 16:58:0010 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 ClosureAt 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.Reactor Protection System
Main Steam Isolation Valve
Feedwater
Primary containment
ENS 5316210 January 2018 15:28:0010 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 SignalAt 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.Reactor Protection System
Main Steam Isolation Valve
Feedwater
Primary containment
High Pressure Coolant Injection
Reactor Core Isolation Cooling
ENS 5264829 March 2017 23:44:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Scram Initiated During StartupAt 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.Intermediate Range Monitor
Reactor Protection System
Reactor Core Isolation Cooling
Primary containment
05000260/LER-2017-003
ENS 5187822 April 2016 18:59:0010 CFR 50.72(b)(3)(iv)(A), System ActuationPartial Loss of Power Experienced During Electrical Bus TransferAt 1359 CDT on April 22, 2016, Browns Ferry Units 1 & 2 experienced a partial loss of power during the transfer of Shutdown Bus 2 from the alternate power source back to the normal power source. During the transfer, the normal feeder breaker failed to close and provide power to the Shutdown Bus, resulting in an auto actuation of two Emergency Diesel Generators (EDGs). Power to Shutdown Bus 2 was immediately restored using the alternate feeder breaker. The EDGs did not tie to the boards. All systems responded as expected for the loss of power, and both Units 1 & 2 maintained 100% Rx Power. All systems have been restored to a normal lineup, and both Units 1 & 2 remain at 100% Rx Power. This event 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), (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified. The cause of the normal feeder breaker failure is being investigated. There was no impact on Unit 3.Emergency Diesel Generator05000259/LER-2016-001
ENS 5133320 August 2015 15:32:0010 CFR 50.72(b)(3)(iv)(A), System ActuationAutomatic Start of an Emergency Diesel Generator on Low Bus VoltageOn August 20, 2015, at approximately 1032 CDT, during the Residual Heat Removal flow rate test, the 3ED 4kV Shutdown Board received a degraded voltage signal, which automatically started the 3D Emergency Diesel Generator (EDG). The EDG performed its safety function by automatically supplying power to the Shutdown Board. Troubleshooting on the degraded voltage signal is in progress. The remaining 4kV Shutdown Boards and EDGs were unaffected and remain operable. This constitutes an automatic actuation of the EDG and requires an 8-hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A), due to the valid actuation of the EDG, and a 60-day report in accordance with 10 CFR 50.73(a)(2)(iv)(A). The licensee has notified the NRC Resident Inspector.Emergency Diesel Generator
Residual Heat Removal
05000296/LER-2015-005
ENS 5040426 August 2014 22:30:0010 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 ScramAt 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.Main Steam Isolation Valve
Feedwater
Primary Containment Isolation System
High Pressure Coolant Injection
Reactor Core Isolation Cooling
Reactor Protection System
ENS 500906 May 2014 13:30:0010 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.

  • * * UPDATE AT 1302 EDT ON 05/09/14 FROM TODD BOHANAN TO DONG PARK * * *

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).

Main Steam Isolation Valve
Feedwater
Primary Containment Isolation System
Reactor Protection System
05000296/LER-2014-002
ENS 4992819 March 2014 03:52:0010 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 LevelAt 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.Main Turbine
Main Steam Isolation Valve
Feedwater
Primary Containment Isolation System
High Pressure Coolant Injection
Reactor Core Isolation Cooling
Reactor Protection System
ENS 4882919 March 2013 09:02:0010 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 VacuumAt 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
ENS 4878225 February 2013 19:13:0010 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 VacuumAt 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.Reactor Protection System
Feedwater
Main Steam Isolation Valve
Reactor Core Isolation Cooling
High Pressure Coolant Injection
Residual Heat Removal
05000296/LER-2013-003
ENS 4890411 February 2013 11:13:0010 CFR 50.72(b)(3)(iv)(A), System ActuationManual Initiation of Reactor Core Isolation Cooling System

On February 11, 2013, at 0613 hours (CDT), the Reactor Core Isolation Cooling (RCIC) system was manually started during a planned Unit 3 reactor shutdown. A Reactor Feedwater recirculation piping separation resulted in the loss of condenser vacuum and subsequent unavailability of the Main Turbine Bypass Valves. The RCIC system was manually started at 9.2" of condenser vacuum in order to control reactor water level in anticipation of loss of Reactor Feedwater Pumps (RFPs) which occurs at 7" of condenser vacuum. Safety Relief Valves (SRVs) were manually operated to maintain reactor pressure. The reactor water level was controlled in the normal band by RCIC, and Reactor Pressure was controlled with a combination of Reactor Core Isolation Cooling (RCIC) system and SRV manual operation. All systems operated as designed and Reactor water level was maintained in the prescribed band. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. RCIC operation was secured at 1449 (CDT) on 2/11/2013.

This event is reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A). During a review of operating logs it was identified that this event met reporting requirements and had not been reported. Therefore, this report does not comply with the 8 hour requirement. This condition has been entered into the corrective action program. Additionally, an LER is required within 60 days per 10CFR50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.

Reactor Core Isolation Cooling
Feedwater
05000296/LER-2013-002
ENS 4862322 December 2012 17:52:0010 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 SystemOn 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.Reactor Protection System
Primary Containment Isolation System
Control Room Emergency Ventilation
Main Steam Isolation Valve
Reactor Core Isolation Cooling
High Pressure Coolant Injection
ENS 4797229 May 2012 08:31:0010 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 SignalOn 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.Main Steam Isolation Valve
Reactor Core Isolation Cooling
Primary containment
Feedwater
Reactor Protection System
Main Transformer
ENS 4795524 May 2012 11:39:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Reactor Scram During StartupAt 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.Intermediate Range Monitor
Reactor Protection System
Reactor Core Isolation Cooling
Primary containment
05000296/LER-2012-004
ENS 4794222 May 2012 07:49:0010 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.

Reactor Protection System
Reactor Core Isolation Cooling
Primary Containment Isolation System
High Pressure Coolant Injection
ENS 4729928 September 2011 09:14:0010 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 TripAt 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.Reactor Protection System
Reactor Core Isolation Cooling
Primary containment
Feedwater
ENS 471306 August 2011 11:17:0010 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Initiation of Technical Specification Required Shutdown

On August 6, 2011, Reactor Protection System (RPS) power supply 1B failed resulting in a partial loss of power to Primary Containment Isolation System (PCIS) groups and an invalid actuation of those PCIS groups. PCIS groups 1 and 2 received partial isolation signals with no subsequent system isolations, as designed. PCIS group 3, 6, and 8 received partial isolation signals with resulting system isolations, also as designed. The combination of loss of RPS 1B and PCIS group 6 isolation resulted in the isolation of the Drywell Floor Drain Sump and the Drywell Continuous Atmospheric Monitor for both particulate and gaseous activity. Thus, both means of automatic monitoring of Reactor Coolant System leakage became inoperable. Unit 1 entered Technical Specification Limiting Condition for Operation (LCO) 3.4.5.D (all required leakage detection systems inoperable) and immediately entered LCO 3.0.3 as required. At the time of occurrence, RPS 1A was being supplied from its alternate source for scheduled maintenance. Thus, the alternate source was not available to RPS 1B. Unit 1 entered LCO 3.0.3 at 0524 (CDT), 'Initiate actions within one hour to place the unit in MODE 2 within 10 hours; MODE 3 within 13 hours; and MODE 4 within 37 hours.' At 0617, Unit 1 began reducing reactor power to comply with LCO 3.0.3. This event requires a 4 hour report IAW 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The PCIS isolations which occurred at 0524 CDT are also reportable within 8 hours per 10CFR 50.72(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)(2), 'General Containment Isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs)), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The event time for the PCIS isolations is 0524 CDT. The NRC resident inspector has been notified. Service Request 412927 was initiated in the Corrective Action Program.

  • * * UPDATE ON 08/06/2011 AT 1350 EDT FROM WILLIAM BAKER TO ERIC SIMPSON * * *

Browns Ferry restored power to the 1B Reactor Protection System power supply at 1208 CDT, reset all isolations and exited LCO 3.0.3. The licensee plans to return the unit to full power. The licensee notified the NRC Resident Inspector. Notified R2DO (Binoy Desai).

Reactor Protection System
Primary Containment Isolation System
Reactor Coolant System
Main Steam Isolation Valve
ENS 4687222 May 2011 22:37:0010 CFR 50.72(b)(3)(iv)(A), System ActuationValid Reactor Protection System (Rps) Actuation Signal - Scram Discharge Volume (Sdv) High Water LevelAt 1737 CDT on 05/22/2011, with BFN U3 (Browns Ferry Unit 3) in Mode 4, a valid RPS actuation signal was received by both channels of the RPS due to Scram Discharge Volume (SDV) High Water Level. At 1735, while performing IRM range 6 to 7 correlation, Instrument Maintenance technicians were measuring high voltage on IRM 'G' while reconnecting a high voltage cable. A spike occurred on IRM's 'C' and 'D' causing a full Reactor Scram. This IRM (Intermediate Range Monitor) Scram was not a valid actuation, the safety function had already been completed, and is not reportable. At 1737, after diagnosing the cause of the IRM Scram, operators reset the Scram signal and received a valid RPS Scram signal due to SDV High Water Level. Investigation is ongoing. This condition is reportable under 10CFR50.72(b)(3)(iv)(A) - A valid actuation of any of the systems named in 50.72(b)(3)(iv)(B). This is also reportable as 60 day written report IAW 10CFR 50.73(a)(2)(iv)(A). This event was entered into the licensee's Corrective Action Program as SR# 373366. There are no compensatory measures or LCO's in effect for this event, and all EDG's and offsite power lines are operable. There was no increase in plant risk as a result of this event. The NRC Resident Inspector has been notified.Reactor Protection System
ENS 468052 May 2011 11:26:0010 CFR 50.72(b)(3)(iv)(A), System ActuationSpecified System Actuation Due to a Emergency Diesel Generator Output Breaker TripAt 0626 (CDT) on 05/2/2011, with BFN U1 in Mode 4, Browns Ferry Nuclear Plant, received an 'A' Emergency Diesel Generator output breaker trip for unknown reasons that de-energized the 'A' 4kV Shutdown Board, resulting in a loss of power to RPS 'A' and a subsequent half scram, 'B' CREV and 'B/C' SGT automatic initiation, and Primary Containment isolation for Groups 2, 3, 6 and 8. The PCIS isolation (Group 2) also caused a loss of Shutdown Cooling on U1 which was restored at 0723 following restoration of power to the 'A' 4kV Shutdown Board from offsite 161kV power. The general containment isolation signals affecting containment isolation valves in more than one system is reportable as an 8 hour notification to the NRC IAW 10CFR50.72(b)(3)(iv), as any event or condition that results in a 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. This is also reportable as 60 day written report IAW 10CFR50.73(a)(2)(iv). The NRC Resident (Inspector) has been notified of this event. This event was entered into the licensee's Corrective Action Program as PER# 362340.Emergency Diesel Generator
Primary containment
Shutdown Cooling
ENS 4680129 April 2011 04:38:0010 CFR 50.72(b)(3)(iv)(A), System ActuationPrimary Containment Isolation System Actuation Due to Loss of Power from a Diesel GeneratorAt 2338 (CDT) on 04/28/2011, with Browns Ferry Nuclear Unit 1 and Unit 2 in Mode 4, Browns Ferry Nuclear Plant, performed a shutdown of the Unit 1/2 Emergency Diesel Generator 'C,' due to an oil leak coming from its governor causing voltage and frequency fluctuations. Following securing of the Unit 1/2 Diesel Generator 'C,' the 4kV Shutdown Board 'C,' which was being powered by DG 'C,' de-energized. This resulted in a loss of power to the 1B RPS, causing a Primary Containment Isolation System (PCIS) actuation and the automatic initiation of the three trains of Standby Gas and 1 train of CREV (Control Room Emergency Ventilation System). The PCIS isolation (Group 2) also caused a loss of Shutdown Cooling on Unit 1 which was restored at 0025 (CDT) 04/29/2011 . In addition, the loss of power to the 4kV Shutdown Board 'C.' also caused the loss of 2B RHR Pump, leading to a momentary suspension of shutdown cooling to Unit 2. Shutdown cooling was immediately restored to Unit 2 using the 2D RHR Pump at 2342 (CDT). The general containment isolation signals affecting containment isolation valves in more than one system is reportable as an 8 hour notification to the NRC IAW 10CFR50.72(b)(3)(iv)(A), as 'Any event or condition that results in a 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.' There were no new Technical Specification LCO's entered as a result of this event. This is also reportable as 60 day written report lAW 10CFR50.73(a)(2)(iv). The NRC Resident (Inspector) has been notified of this event. This event was entered into the licensee's Corrective Action Program as SR# 361382.Primary Containment Isolation System
Emergency Diesel Generator
Shutdown Cooling
ENS 4679327 April 2011 22:01:0010 CFR 50.72(a)(1)(i), Emergency Class Declaration
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
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.

  • * * UPDATE FROM BILL BAKER TO HOWIE CROUCH AT 1942 EDT ON 4/27/11 * * *

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.

  • * * UPDATE FROM BILL BAKER TO S. SANDIN AT 2153 EDT ON 4/27/11 * * *

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.

  • * * UPDATE FROM BILL BAKER TO HOWIE CROUCH AT 2303 EDT ON 4/27/11 * * *

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).

  • * * UPDATE FROM BILL BUTLER TO HOWIE CROUCH AT 2338 EDT ON 4/27/11 * * *

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.

  • * * UPDATE FROM GIVENS TO HUFFMAN AT 2200 EDT ON 5/2/11 * * *

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).

Emergency Diesel Generator
Reactor Protection System
Primary Containment Isolation System
Primary containment
Main Steam Isolation Valve
High Pressure Coolant Injection
Reactor Core Isolation Cooling
05000259/LER-2011-001
ENS 4651126 December 2010 22:20:0010 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 ExciterOn 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.Primary Containment Isolation System
Reactor Protection System
05000296/LER-2010-004
ENS 459909 June 2010 08:31:0010 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).

Main Steam Isolation Valve
Reactor Protection System
Reactor Core Isolation Cooling
High Pressure Coolant Injection
Feedwater
Intermediate Range Monitor
ENS 4588830 April 2010 21:48:0010 CFR 50.72(b)(3)(iv)(A), System ActuationLight Socket Short Causes Isolation Systems to Actuate

At 1648 CDT, while the Control Bay AUO (Auxiliary Unit Operator) was attempting to change a light bulb for 1-IL-99-1AA, 1A (Reactor Protection System) RPS (Motor Generator) MG Set available light, the light socket shorted causing the loss of 1A RPS. The loss of 1A RPS resulted in Groups 2, 3, 6, and 8 primary containment isolations and initiated Standby Gas Treatment and Control Room Emergency Ventilation. All systems responded as designed. The Control Room operators entered the appropriate abnormal operating instruction, 1-AOI-99-1, to restore the affected systems. Operations entered TS LCO 3.3.1.1 conditions A.1 (place channel in Trip in 12 hours) and C.1 (restore RPS trip capability in 1 hr). At 1734 CDT, Operations exited TS LCO 3.3.1.1 upon restoration of 1A RPS per 1-AOI-99-1. This event is reportable within 8 hours per 10CFR 50.72(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) (b. General Containment Isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs)), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM RAY SWAFFORD TO DONG PARK AT 1522 EDT ON 6/29/10 * * *

Retraction of an 8 hour non-emergency notification for invalid Primary Containment Isolation System (PCIS) actuation from a loss of power to the Reactor Protection System (RPS) 1A. Browns Ferry Nuclear Plant (BFN) is retracting the 10 CFR 50.72(b)(3)(iv)(A) eight-hour non-emergency report made on April 30, 2010, at 2131 hours Central Daylight Time. BFN's initial report was categorized as a valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B). The loss of power to RPS 1A resulted in PCIS Groups 2, 3, 6, and 8 primary containment isolations and initiation of Standby Gas Treatment and Control Room Emergency Ventilation. However, plant conditions which require PCIS actuations and system initiations (e.g., low reactor water level, high drywell pressure, abnormal area radiation level, high area temperature) did not exist, therefore, the actuation was invalid. As a consequence, TVA has concluded that this event does not meet the reporting requirements of 10 CFR 50.72. The event is reportable under 10 CFR 50.73(a)(2)(iv)(A). A 60-day phone call will be made in accordance with 10 CFR 50.73(a)(1). (See EN #46054) TVA's evaluation of this event is documented in the Corrective Action Program (PER 227662). TVA has notified the NRC Resident Inspector. Notified R2DO (Desai).

Primary containment
Control Room Emergency Ventilation
Main Steam Isolation Valve
Primary Containment Isolation System
Reactor Protection System
ENS 4539130 September 2009 04:23:0010 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.

  • * * UPDATE FROM MIKE HUNTER TO JOE O'HARA AT 1508 ON 9/30/09 * * *

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
Core Spray
Reactor Pressure Vessel
05000260/LER-2009-008
ENS 4529024 August 2009 23:50:0010 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Condensate Booster Pump Trip Resulting in a Manual Reactor ScramOn 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.Secondary containment
Feedwater
Reactor Protection System
ENS 4486018 February 2009 09:51:0010 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.

  • * * * UPDATE FROM RICKY GIVENS TO JOHN KNOKE AT 1828 0N 02/21/09 * * * *

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)

Reactor Protection System
Reactor Core Isolation Cooling
Feedwater
ENS 4485416 February 2009 11:13:0010 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 WaterAt 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 Water
ENS 445405 October 2008 03:08:0010 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 GeneratorOn 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.Reactor Core Isolation Cooling
Feedwater
ENS 441875 May 2008 08:32:0010 CFR 50.72(b)(3)(iv)(A), System ActuationLoss of Power to Esf Buses During Transfer of Offsite Power SourceWith Unit 3 in a refueling outage and in mode 5, as part of outage activities 4160V unit board transfers were in progress to return Unit 3 to 500KV offsite source, the 3B 4160V Unit Board failed to transfer. The loss of 3B 4160V Unit Board resulted in loss of power to 3EC and 3ED 4160V shutdown boards, 3B 480V shutdown board, 480V SBGT board, and 3B RPS. 3EC and 3ED diesel generators started and tied to the 4160V shutdown boards as required for loss of power. Unit 3 groups 3 and 6 isolations were received as expected. An unexpected full scram was received. An invalid A IRM actuation occurred (A channel half scram) while RPS 3B bus de-energized resulting in a full scram signal. Previous to this event all control rods were already inserted. This non-emergency event is reportable under 50.72(b)(3)(iv)(A)(h) - Any event or condition that results in valid actuation of emergency ac electrical power systems. UNIT 3 remains in Mode 5, flooded up with fuel pool gates removed and all rods are fully inserted. 3A, 3B and 3C 4160V Unit Boards and all four shutdown boards are energized. Units 1 and 2 were operating at 100% power and were not affected by the incident except for the common refuel zone isolation and temporary loss of control bay ventilation. The EDGs have been restored to standby and the electrical configuration has been returned to normal. The licensee notified the NRC Resident Inspector.05000296/LER-2008-001
ENS 438781 January 2008 03:40:0010 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 SignalOn 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.Reactor Core Isolation Cooling
Feedwater
Reactor Protection System
ENS 4371812 October 2007 13:02:0010 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 TripAt 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 Cooling05000259/LER-2007-009
ENS 436133 September 2007 07:14:0010 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.

  • * * UPDATE FROM DON SMITH TO HUFFMAN AT 1357 EDT ON 9/05/07 * * *

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.

Shutdown Cooling
Feedwater
05000259/LER-2007-008
ENS 4356011 August 2007 22:51:0010 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 SignalOn 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.Primary Containment Isolation System
Reactor Protection System
ENS 434149 June 2007 16:00:0010 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).

  • * * UPDATE FROM BOLAND TO HUFFMAN AT 1850 EDT ON 6/9/07 * * *

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.

  • * * UPDATE FROM TIM GOLDEN TO JOE O'HARA AT 1804 ON 6/15/07 * * *

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).

Primary Containment Isolation System
Reactor Protection System
05000259/LER-2007-005
ENS 4338124 May 2007 07:11:0010 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 LeakOn 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.Main Turbine
Primary containment
ENS 431599 February 2007 18:08:0010 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
Reactor Scram Due to Low Reactor Water LevelAt 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.Feedwater
High Pressure Coolant Injection
Reactor Core Isolation Cooling
Primary containment
Shutdown Cooling
Control Room Emergency Ventilation
Reactor Protection System
Reactor Coolant System
05000296/LER-2007-001
ENS 4309211 January 2007 14:16:0010 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 OpenedUnit 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.Feedwater05000260/LER-2007-001
ENS 4281330 August 2006 03:25:0010 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 LeakOn 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
ENS 4278719 August 2006 16:05:0010 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 PumpsAt 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.Reactor Pressure Vessel
Primary Containment Isolation System
Shutdown Cooling
Control Room Emergency Ventilation
Feedwater
Standby Gas Treatment System
05000296/LER-2006-002
ENS 4210231 October 2005 19:18:0010 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 TurbineAt 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.Main Turbine
Feedwater
Shutdown Cooling
05000296/LER-2005-003
ENS 4199717 September 2005 16:29:0010 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
Control Room Emergency Ventilation
Reactor Water Cleanup
05000296/LER-2005-002
ENS 418965 August 2005 22:08:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge
10 CFR 50.72(b)(3)(iv)(A), System Actuation
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.

  • * * UPDATE PROVIDED BY HUNTER TO JEFF ROTTON AT 1524 ON 08/06/05 * * *

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).

05000260/LER-2005-007
ENS 4159513 April 2005 15:53:0010 CFR 50.72(b)(3)(iv)(A), System ActuationValid Rps Actuation Due to an Invalid Atws Logic ActuationAt 0953 (CDT) on 04/13/2005 with Unit 2 in MODE 4 (and) all Control Rods inserted, a valid RPS Actuation (Reactor SCRAM Signal resulting from the filling of the SCRAM DISCHARGE VOLUME) was received as a result of an invalid ATWS initiation. This occurred during the performance of 2-SR-3.3.5.2.4(FT) RCIC SYSTEM LOGIC FUNCTIONAL TEST. Further review of the test revealed that a Volt/Ohm Meter was attached across contacts in the ATWS LOGIC that conflicted with the RCIC Surveillance and apparently resulted in the ATWS initiation and the filling of the SCRAM DISCHARGE VOLUME. This event is reportable as an 8-hour Non-Emergency Notification along with a 60-day written report in accordance with 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 as described in (1) & (2) below.' The licensee notified the NRC Resident Inspector.05000260/LER-2005-003
ENS 4140411 February 2005 22:29:0010 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 RejectionThe 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.05000296/LER-2005-001
ENS 4121923 November 2004 16:02:0010 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 TripAt 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.05000296/LER-2004-002
ENS 4086211 July 2004 03:35:0010 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 MonitorsOn 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).Intermediate Range Monitor
Reactor Protection System
ENS 408589 July 2004 03:32:0010 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 SignalThe 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.Primary Containment Isolation System
Primary containment
Secondary containment
Main Turbine
Feedwater
Reactor Protection System