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The query [[Category:ENS Notification]] [[Site::Browns Ferry]] [[Scram::+]] was answered by the SMWSQLStore3 in 0.4596 seconds.


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 Entered dateSiteRegionScramReactor typeEvent description
ENS 5392310 March 2019 04:38:00Browns FerryNRC Region 2Automatic ScramAt 2259 CST on 3/9/2019, Browns Ferry Unit-3 received an automatic SCRAM on Main Generator Breaker Failure and Turbine Load Reject. Unit-3 declared a Notification of Unusual Event SU1 for loss of offsite AC power to Unit-3 specific 4kV Shutdown Boards for greater than 15 minutes. Primary Containment Isolation Systems (PCIS) Groups 1, 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required. Main steam relief valves lifted on the initial transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated on low reactor water level. HPCI remains in service for reactor level and pressure control. RCIC is not in service at this time, the station is investigating low flow from the pump. All four Unit-3 Diesel Generators started and loaded as expected. Residual Heat Removal System is in service for suppression pool cooling. 4kV Station Unit Boards have been restored from the 161kV system. Actions are in progress to restore 4kV Shutdown Boards to offsite power. This event is reportable within 1 hour in accordance with 10 CFR 50.72(a)(1)(i) for declaration of the Licensees Emergency Plan. Complete as documented on EN 53922. This event requires a 4 hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A). 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs), (4) ECCS (Emergency Core Cooling System) for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system, (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system, and (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified. As of the event report, the MSIVs were opened and decay heat was being removed via the bypass valves to the condenser.
ENS 5326918 March 2018 16:16:00Browns FerryNRC Region 2Automatic ScramGE-4At 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.
ENS 5316210 January 2018 13:53:00Browns FerryNRC Region 2Automatic ScramGE-4At 0928 CST on January 10, 2018, the Unit 3 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from Turbine Control Valve Emergency Trip System pressure low. The reactor had been operating near 73 percent power for an emergent issue for Turbine Control Valve (TCV) No. 3. With TCV No. 3 out of service and closed, the unit was operating with RPS in a half scram condition. A subsequent failure of the TCV No. 2 sensing line resulted in RPS coincidence logic being met for TCV fast closure SCRAM. The investigation of the TCV No. 2 sensing line failure continues. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Turbine Bypass Valves controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical except when the actuation results from and is part of a preplanned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident inspector has been notified.
ENS 5264829 March 2017 23:36:00Browns FerryNRC Region 2Manual ScramGE-4At 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.
ENS 5040426 August 2014 21:24:00Browns FerryNRC Region 2Automatic ScramGE-4At 1730 CDT on August 26, 2014, Browns Ferry Unit 1 experienced a turbine trip resulting in an automatic reactor scram. The cause of the turbine trip was a control valve fast closure signal that was generated by a turbine trip on generator neutral over voltage signal. The Main Steam Isolation Valves (MSIVs) remained open with the main turbine bypass valves controlling reactor pressure. The Reactor Feedwater Pumps are in service to control reactor water level. Primary Containment Isolation Systems (PCIS) Groups 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required with the exception of Standby Gas Treatment (SBGT) train A, which is under a clearance for planned maintenance. Neither High Pressure Coolant Injection (HPCI) nor Reactor Core Isolation Cooling (RCIC) initiation signals were received. Initially, three Main Steam Relief Valves (MSRVs) opened to control the pressure surge and subsequently reclosed. This event requires a 4 hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including reactor scram or reactor trip, and (2) General containment Isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).' The NRC Resident Inspector has been notified. Service Request 926468 was initiated in the Corrective Action Program. The plant is in its normal shutdown electrical lineup. The licensee is investigating the cause of the generator neutral overvoltage signal. There was no impact on units 2 and 3.
ENS 500906 May 2014 13:27:00Browns FerryNRC Region 2Automatic ScramGE-4

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).

ENS 4882919 March 2013 08:37:00Browns FerryNRC Region 2Manual ScramGE-4At 0402 (CDT) on 03/19/2013, the Unit 1 reactor was manually scrammed due to lowering main condenser vacuum. The cause of the loss of vacuum was a significant leak on the 1C feedwater heater level control line. The leak appeared as a steam/water leak near the penetration to the main condenser. As extraction steam was isolated, condenser vacuum deteriorated and was approaching the turbine trip setpoint, at which time the reactor was manually scrammed. Condenser vacuum recovered following the scram. MSIVs (Main Steam Isolation Valves) are open, main turbine bypass valves are controlling reactor pressure and reactor feedwater pumps are being used to control reactor water level. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated as required. In response to the scram, all plant equipment responded as designed. The reactor had been operating near 95% power for several hours due to the 1C3 heater isolating at 2334 (hrs. CDT) on 3/18/13. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B) `any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR50.73(a)(2)(iv)(A). All rods inserted into the core during the scram. No safety relief valves lifted during the transient. The plant is in its normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector.
ENS 4878225 February 2013 17:49:00Browns FerryNRC Region 2Automatic ScramGE-4At 1313 (CST) on 02/25/2013, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip. Preliminary indications show the turbine tripped on low condenser vacuum. Cause of loss of condenser vacuum has been identified as Reactor Feedwater recirculation piping separation. Main Steam Isolation Valves (MSIVs) were manually closed to isolate the leak. None of the Safety Relief Valves (SRVs) automatically cycled during the transient, and one Safety Relief Valve (SRV) was manually operated to maintain Reactor Pressure due to the Main Turbine Bypass Valves unavailability because of loss of condenser vacuum. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC), reactor water level initiation set points were reached. Reactor water level is being controlled by the RCIC system and Reactor Pressure is being controlled with the High Pressure Coolant Injection (HPCI) system. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated, with the exception of one valve in Group 6. Drywell Continuous Air Monitor (CAM) Inboard Return Isolation Valve 3-FSV-90-257 did not have indication following isolation signal and was not able to be verified locally. Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM was manually isolated at 1422 (CST) with positive indication of isolation, and isolation valves deactivated at time 1514 (CST) to satisfy TS LCO 3.6.1.3 required actions. This event is reportable within 4 hours per 10CFR50.72(b)(2)( iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). At 1415 (CST), Suppression Pool Water level exceeded -1 inch due to operation with HPCI in pressure control mode, and required entry into TS LCO 3.6.2.2 condition A to restore level within 2 hours. Efforts are being made to lower suppression pool water level within limits. At 1615 (CST), water level remains above -1 inch requiring entry into TS LCO 3.6.2.2 condition B requiring action to be in MODE 3 in 12 hours and MODE 4 within 36 hours. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. All control rods fully inserted and electrical offsite power is in a normal shutdown configuration. Residual Heat Removal is aligned for suppression pool cooling. There was no impact on either Unit 1 or 2.
ENS 4862322 December 2012 16:39:00Browns FerryNRC Region 2Automatic ScramGE-4On 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.
ENS 4797229 May 2012 07:22:00Browns FerryNRC Region 2Automatic ScramGE-4On 5/29/2012 at 0331 (CDT) the Unit 3 reactor scrammed due to turbine control valve fast closure initiated by a load reject signal on the Main Generator. The cause of the load reject signal is Main Transformer differential relay 387T. Reactor power at the time of the SCRAM was approximately 75%. All systems responded as expected to the load reject signal. Main Steam Isolation Valves remained open and reactor pressure is being controlled on the Main Turbine Bypass Valves. No Main Steam Relief Valves lifted during the transient. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation setpoints were reached. Primary Containment Isolation Signals (PCIS) Groups 2, 3, 6 and 8 were received. The lowest reactor water level observed was -41 inches. Reactor water level was restored to and is being controlled by the Feedwater system in the normal band. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B), 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR50.73(a)(2)(iv)(A). This event is documented in the station corrective action program on SR# 557947. The NRC Resident Inspector has been notified. All control rods inserted into the reactor core. Electrical power is being back fed from offsite power through the 161 KV feeder line. The reactor is being cooled down within the Technical Specification rates.
ENS 4795524 May 2012 11:10:00Browns FerryNRC Region 2Manual ScramGE-4At 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.
ENS 4794222 May 2012 07:38:00Browns FerryNRC Region 2Automatic ScramGE-4

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.

ENS 4785319 April 2012 20:58:00Browns FerryNRC Region 2Manual ScramGE-4On 04/19/12 at 1430 while performing 1-SR-3.5.1.7, HPCI (High Pressure Coolant Injection) Main & Booster Pump Set developed head & flow rate at rated reactor pressure. The HPCI turbine failed to trip using the manual trip pushbutton. This manual trip pushbutton should have caused the 1-FCV-73-18, HPCI TURBINE STOP VALVE, to go closed. HPCI was secured by taking the 1-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, to close. The 1-FCV-73-18, HPCI TURBINE STOP VALVE, also failed to go closed locally using the 1-XCV-73-18, HPCI TURBINE MECHANICAL TRIP, nor did it go closed when the auxiliary oil pump was secured. With the 1-FCV-73-18, HPCI TURBINE STOP VALVE, open, the HPCI ramp generator is no longer in the circuit therefore, should an initiation occur and cause the 1-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, to open there is the potential for the HPCI turbine to over speed. Therefore, HPCI was isolated using 1-FCV-73-3, HPCI STEAM LINE OUTBD ISOL VALVE. This incident is reportable as an 8-hour ENS notification under 10CFR 50,72 (b)(3)(v) as 'any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.' It also requires a 60 day written report in accordance with 10CFR 50.73(a)(2)(vii). The NRC Resident Inspector has been notified.
ENS 476518 February 2012 19:31:00Browns FerryNRC Region 2Manual ScramGE-4During BFNP NFPA 805 transition review, it was determined in the vent of an Appendix-R fire, the Reactor Protection System (RPS) function could be rendered not functional. The current Appendix R Safe Shutdown Analysis states: "The safe shutdown function of the Reactor Protection System (RPS) is to initiate reactor scram through actuation of the control rod drives. The RPS includes the RPS motor-generator power supplies and associated control and indicating devices, sensors, relays, bypass circuitry, and switches that initiate rapid insertion of control rods (scram) to shutdown the reactor. The RPS utilizes a fail-safe design so that device failures or a loss of power will result in control rod insertion. The scram function will remain available despite any fire-induced spurious signals that may be generated due to the effects of a postulated fire in any fire area. This system is expected to perform its function automatically, however credit is taken only for manual scram. No additional analysis is needed to ensure the availability of reactor scram in the even of a fire. Due to lack of physical separation with 120 volt AC lighting circuitry, the RPS system potentially could remain energized due to a postulated hot short circuit during a fire which could potentially prevent the control rods from inserting. Therefore, the fail-safe design of the RPS system would not be maintained. Compensatory actions in the form of fire watches to mitigate this condition are in place in accordance with the BFNP Fire Protection Report. This event is reportable as an 8-hour notification to the NRC in accordance with 10CFR50.73(a)(2)(ii)(B). The NRC Resident Inspector has been notified of this event. This event was entered into the licensee's Corrective Action Program as PER 503304.
ENS 4729928 September 2011 08:26:00Browns FerryNRC Region 2Automatic ScramGE-4At 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.
ENS 4651126 December 2010 20:53:00Browns FerryNRC Region 2Manual ScramGE-4On 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.
ENS 4539130 September 2009 04:09:00Browns FerryNRC Region 2Automatic Scram
Manual Scram
GE-4

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).

ENS 4529024 August 2009 23:38:00Browns FerryNRC Region 2Automatic Scram
Manual Scram
GE-4On 8/24/09, at 18:50 Unit 3 was manually scrammed due to loss of 2 of the 3 Condensate Booster Pumps due to low pump suction pressure. The cause for the Condensate Booster Pump low suction pressures is unknown at this time, but is under investigation. After the reactor was scrammed manually, reactor water level lowered below the automatic scram set point (+2 inches) and below the automatic start for HPCI and RCIC (-45 inches). All expected Primary and Secondary Containment isolation valves operated as required, isolation groups 2,3,6 and 8 were actuated. Both reactor recirculation pumps tripped due to the low reactor water level. HPCI and RCIC actuated as expected to restore reactor water level. Reactor pressure control was maintained on the turbine bypass valves, and no Main Steam Relief Valves (MSRVs) were opened as a result of the transient. At this time the unit is stable in mode 3. Reactor water level is being controlled using one Reactor Feedwater pump, HPCI and RCIC have been returned to standby readiness. The 3B Reactor Recirculation Pump has been returned to service. Reactor pressure is being automatically maintained by the main turbine bypass valves. This event is reportable as a 4 hour non-emergency report due to 10CFR 50.72(b)(2)(iv)(A) and (B) (ECCS discharge to the reactor and Reactor Protection System (RPS) actuation) and as an 8 hour non-emergency report due to 10CFR50.72(b)(3)(iv)(A) (specified system actuations). All rods fully inserted on the SCRAM. The plant is in its normal shutdown lineup. The licensee notified the NRC Resident Inspector.
ENS 4512411 June 2009 18:24:00Browns FerryNRC Region 2Manual ScramGE-4

At noon on 06/11/2009 Browns Ferry Unit 2 experienced a rise in drywell leakage during reactor startup. The 4 hour unidentified leak rate from 0800 to 1200 on 6/10/2009 was 0 GPM, while the 4 hour unidentified leak rate from 0800 to 1200 on 6/11/2009 was 3.88 GPM. This increase in leak rate exceeded the allowable limit of a 2 GPM increase in unidentified leakage in a 24 hour period per Technical Specification 3.4.4. At 1555 (CDT) on 06/11/2009, Unit 2 inserted a manual reactor SCRAM to comply with Technical Specification 3.4.4 Condition C to be in Mode 3 in 12 hours and Mode 4 within 36 hours. All systems responded as required except that when reset of the SCRAM was attempted after all control rods had inserted, a portion of the 'B' RPS channel (B2/B3) failed to reset. The 'A' RPS channel was then actuated by a spike on the Intermediate Range Monitors while they were being driven into the core resulting in a full RPS SCRAM actuation. The licensee notified the NRC Resident Inspector." Unit 2 is currently removing decay heat using normal feedwater with bypass steam to the main condenser. The electrical system is in a normal shutdown configuration. The licensee has not identified the source of the unidentified leakage and anticipates taking Unit 2 to Cold Shutdown (Mode 4).

  • * * UPDATE AT 1730 EDT ON 6/15/09 FROM HUNTER TO SANDIN * * *

Plant personnel entered the drywell after the reactor was shutdown. The resulting investigation revealed two sources of increased drywell leakage. One source of leakage was from 2-CKV-10-511 and 2-CKV-10-526, SRV Tailpipe Vacuum Breakers, which were apparently damaged by leakage past the pilot valve for 2-PCV-001-0023, Main Steam Safety Relief Valve. The combination of damaged vacuum breakers and a leaking pilot valve resulted in steam leakage into the drywell atmosphere. The other source of leakage was noted through the packing for 2-DRV-10-505, RPV Drain to RWCU. The vacuum breakers, pilot valve, and drain valve have been repaired. The 'B' RPS contactors were repaired and the scram reset. The NRC Resident Inspector has been notified." R2DO (McCoy) notified.

ENS 4485416 February 2009 09:45:00Browns FerryNRC Region 2Manual ScramGE-4At 0513 on 2/16/09, the Unit 2 reactor was manually scrammed in accordance with alarm response procedure 2-ARP-9-8A 'TURBINE TRIP TIMER INITIATED'. Other associated alarms and indications both locally and in the Main Control Room indicated a failure of the stator cooling water system. The exact cause of the failure is still being investigated. All systems responded as expected to the insertion of the manual scram. No ECCS injection was initiated or required, and all expected containment isolation and initiation signals were received. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). All rods inserted fully into the reactor. The electrical power system is in a normal shut down configuration. Decay heat removal is through the main condenser via the turbine bypass valves. There is no impact on Units 1 and 3. The NRC resident inspector has been notified.
ENS 438781 January 2008 01:58:00Browns FerryNRC Region 2Automatic ScramGE-4On 12/31/07 at 2140 the Unit 3 reactor scrammed due to turbine generator load reject signal on the Main Generator. The cause of the load reject signal is unknown and the investigation is continuing. All systems responded as expected to the load reject signal. Six Main Steam Relief valves (MSRVs) opened momentarily and then reclosed. Subsequently, reactor pressure was automatically controlled by the Main Turbine Bypass valves. No Emergency Core Cooling System (ECCS), nor Reactor Core Isolation Cooling (RCIC) reactor water level initiation setpoints were reached, and reactor water level is being automatically controlled by the Feedwater system. This report is being made as required by 10CFR 50.72(b)(2) due to the actuation of the Reactor protection System. Refer to BFN PER number 135878. All control rods fully inserted into the core, and all safety systems are operable. PCIS group isolations were received for groups 2, 3, 6, and 8. There were no grid abnormalities at the time of the load reject, and the event had no effect on Unit 1 or 2. The licensee notified the NRC Resident Inspector.
ENS 4371812 October 2007 11:07:00Browns FerryNRC Region 2Automatic ScramGE-4At 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.
ENS 436133 September 2007 05:05:00Browns FerryNRC Region 2Manual ScramGE-4

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.

ENS 4356011 August 2007 21:25:00Browns FerryNRC Region 2Automatic ScramGE-4On 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.
ENS 434149 June 2007 15:53:00Browns FerryNRC Region 2Automatic ScramGE-4

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).

ENS 4338124 May 2007 04:42:00Browns FerryNRC Region 2Manual ScramGE-4On 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.
ENS 431599 February 2007 16:32:00Browns FerryNRC Region 2Automatic ScramGE-4At 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.
ENS 428785 October 2006 14:18:00Browns FerryNRC Region 2Manual ScramGE-4This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of multiple main steam isolation valves. On August 19, 2006, at 2002 hours CDT, with Unit 3 shutdown in Mode 3 following an earlier manual scram of Unit 3 (reference Event Number 42787), the EHC System along with the turbine bypass valves was being used to cool down/depressurize the reactor. The EHC system is used to supervise the turbine control valves and turbine bypass valves to control reactor pressure. The EHC system was subsequently removed from service for the repair of a previously identified system fluid leak. Following the shutdown of EHC system, reactor pressure and temperature were allowed to slowly increase. At 2240 hours CDT upon completion of the leak repair, operations personnel placed an EHC pump back in service in accordance with the system operating instruction to support post-maintenance testing activities. However, when the EHC system was returned to service, the pressure control set point was lower than the reactor pressure. The significance of the delta between the EHC setpoint and the actual reactor pressure was not recognized by operations. Because the actual reactor pressure was higher than the existing control set point, the EHC system responded by opening turbine bypass valves to lower the reactor pressure. Operations personnel observed the bypass valve response, identified the cause, and raised the pressure control set point. This action caused the bypass valves to rapidly close. The abrupt cessation of steam flow caused by the rapid closure of the bypass valves initiated a reactor pressure transient that affected the reactor water level instrumentation. The affected level instruments' output signals exhibited a ringing effect of a magnitude sufficient to reach the low level set point for primary containment isolation system (PCIS) Group 1 actuation. The Group 1 isolation logic actuated in accordance with its design, and the main steam isolation valves and the main steam line drain valves automatically closed. Designed time delays in other logic circuits affected by these water level signals prevented additional equipment actuation during this event. Actual reactor water level did not change, remaining within the normal level band; therefore, the isolation signal is considered invalid. All equipment responded in accordance with the plant design. Upon verification that no actual water level anomaly existed and that the transient instrumentation response had stabilized, the affected PCIS logic was reset, and equipment was realigned as appropriate. There were no safety consequences or impacts on the health and safety of the public. The event was entered into TVA's corrective action program for evaluation and resolution. The NRC senior resident inspector has been notified of this report. Reference corrective action document PER 109118.
ENS 4281330 August 2006 02:52:00Browns FerryNRC Region 2Manual ScramGE-4On 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.
ENS 4278719 August 2006 15:15:00Browns FerryNRC Region 2Manual ScramGE-4At 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.
ENS 4210231 October 2005 17:37:00Browns FerryNRC Region 2Automatic ScramGE-4At 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.
ENS 4140411 February 2005 20:29:00Browns FerryNRC Region 2Automatic ScramGE-4The following text was obtained from the licensee via facsimile: At 1629 (hrs. CST) on 02/11/05, Unit 3 reactor scrammed from 100% power when the output breaker tripped causing a load reject. The breaker tripped due to a corresponding switchyard breaker 5268 tripping when a PK block was re-installed before the trip cutout (TCO) switches were placed in TCO. All rods inserted. Unit 2 was also at 100% power and was unaffected by this event. Water level lowered to +1" as expected and was recovered by normal feed water flow. All expected PCIS (Primary Containment Isolation System) isolations, Group 2 (RHR S/D (Residual Heat Removal) cooling), Group 3 (RWCU (Reactor Water Clean Up)), Group 6 (ventilation), and group 8 (TIP (Transverse Incore Probe)) were received along with the auto start of CREV (Control Room Emergency Ventilation) and the 3 SGT (Standby Gas Treatment) trains. Four MSRV's (Main Steam Safety Relief Valves) lifted momentarily to stabilize reactor pressure. This event is reportable as a 4-hour and 8-hour Non-Emergency Notification along with a 60-day written report in accordance with 10CFR50.72(b)(2)(iv)(B), 10CFR50.72(b)(3)(iv)(A) and 10CFR50.73(a)(2)(iv)(A) as 'Any event or condition that results in a valid actuation of RPS (Reactor Protection System) and PCIS .. The plant was performing restoration from switchyard maintenance at the time of the scram. All safety relief valves that lifted properly re-seated. Decay heat is being removed via the steam bypass valves to condenser. The electrical grid is stable. The licensee has notified the NRC Resident Inspector.
ENS 4121923 November 2004 14:20:00Browns FerryNRC Region 2Automatic ScramGE-4At 1002 on 11/23/2004 with Unit 3 @ 100% power, a full reactor scram signal (RPS) was received due to a Turbine Trip. Unit 2 was also @ 100% power and was unaffected by the event. Reactor water level lowered to approximately -30 (inches) as expected and was recovered with normal feed water flow. All expected PCIS ISOLATIONS, Group 2 (RHR S/D Cooling), Group 3 (RWCU), Group 6 (Ventilation) & Group 8 (TIP) were received along with the auto start of CREVS and the three SBGT Trains. One PCIS Scram Discharge Volume Drain (85-37F) failed to close but the flowpath did isolate. The Unit remains Shutdown and an investigation is underway to determine the cause of the Turbine Trip and resulting Reactor Scram. This event is reportable as a 4-hour & 8-hour Non-Emergency Notification along with a 60 day written Report in accordance with 10CFR50.72(b)(2)(iv)(B), 10CFR50.72(b)(3)(iv)(A) & 10CFR50.73(a)(2)(iv)(A) as 'Any event or condition that results in valid actuation of RPS & PCIS . . . The licensee informed the NRC Resident Inspector.
ENS 4086211 July 2004 01:31:00Browns FerryNRC Region 2Automatic ScramGE-4On 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).